ML20125D632

AP600 Tier 1 Matl,Insps,Test,Analysis & Acceptance Criteria

ML20125D632
Person / Time
Site:	05200003
Issue date:	12/15/1992
From:	Antolovich D, Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19303F098	List: ML20125D612 ML20125D617 ML20125D621 ML20125D632
References
GW-GL-030, GW-GL-030-R01, GW-GL-30, GW-GL-30-R1, NUDOCS 9212150334
Download: ML20125D632 (179)

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l APM) DETAll.E') DESIGN AND DESIGN i CERTIFICATION PROGRAM CONTROI.I.ED DOCUMENT DISTRilltITION RECORD SHEET Project APM) Procram Document Title APm) TIER 1 MATERI AL - Plant Description inspections. Tests,' Analvscs, and Acceptance Criteria Document identification Number GW G1,030 - REV,1 Controlled Ccpy 1 40 Uncontrolled Copy Assigned To Nuclear Reculatory Comminion Revisions of this Document will be distributed to holders on Controlled copies. The Document holder of a Controlled copy in responsible for entering all changes in his copy of the Document and for destroying all superseded page/ document. This Document is the property of Westinghouse NNIT).

Please return this Document to the Records Coordinator, at the address below, when it is no longer needed or in the event of position reassignment.

Susan L Keaney Program Control & Contract Administration Energy Center East - Bay 462 Westinghouse Electric Corporation P.O. Box 355 g  ;

Pittsburgh, Pennsylvania 15230 l

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l 9212150334 921215 PDR ADOCK 0S200003 A PDR

AP600 DOCUMENT COVER SHEET Form 58202D(5/92)[WPxxxx:10) AP600 DD USE ONLY Pages Attached a w nu AP600 DOCUMENT NO. REVISION NO. DATED CONTROLLED COPY NUMBER:

GW GL 030 1 Dec. 15, 1992 ASSIGNED TO:

ALTERNATE DOCUMENT NUMBER: ATTACHMENTS DESIGN AGENT ORGANIZATION:

PROJECT: AP600 TITLE: AP600 TIER 1 MATERIAL o Plant Description o inspections, Tests, Analysis and Acceptance criteria WORK BREAKDOWN #: 3.2.5 This section incorporates the lobowing desgn changes DCP 4/Rev.:

0 (C) WESTINGHOUSE ELECTRIC CORPORATION 1991 A ucENsE IS RESERVED TO THE U.S. GOVEFWMENT UNDER CONTRACT DE AC03@0SF18895.

O WESTINGHOUSE PROPRIETARY CLASS 2 THIS DOCUMENT CONT AINS NFORMATON PROPRIETARY TO WESTINo>OUsE ELECTRIC CORPORATON; IT IS SUDMITIED IN CONFIDENCE AND l$ TO BE USED d SOLctY FOR THE PURPOSE FoR WHiCH rf 16 FURNISHED AND RETURNED UPON REOUEST. THis DOCUMENT AND BUCH INFORMATON 18 NOT TO BE REPRODuctD, TRANsurtTED, DISCLOSED OR USED OTHERWISE N WHOLE OR IN PART WITtOUT PRIOR WRITTEN AUT)OR12ATON OF WESTWGHOUSE ELECTRIC CORPORATION, ENERoY SYSTEMS DOSNEss UNIT, SUBJECT TO THE LEGENDS CONTAWED HEREOF. '

GOVERNMENT LIMITED RIGHTS:

(A) THE$t DATA ARE SupuitTED WiTH LluiTED RioHTS UNDER GOVERNMENT CONTMCT NO. DE-ACoawSF18495. THESE DATA MAY BE REPRODUCED -

AND USED BY THE GOVERNMENT WITH THE EXPRESS UMITATON THAT THEY WILL NOT, WITHOUT WRITTEN P8P6'1SSON OF THE CONTRACTOR, BE USED FOR PURPOSES OF MANUFACTURER NOR DISCLOSED OUTSOE THE GOVERNMENT: EXCEPT THAT THE GodHrdMtNT MAY DISCLOSE THESE DATA OUTSOE THE GOVERNMENT FOR THE FOLLOWINO PURPOSES, IF ANY, PROVIDED THAT THE GOVERNMENT MAKES SU,# DISCLOSURE SUEUECT TO PROHIBITION AGANST FURTHER USE AND DISCLOSURE (1) THl$ 'PFCPRIETARY DATA' MAY BE DISCLOSED FOR EVALUATION PURPOSES UNDER THE RESTRICTIONS AROVE. .

i (11) THE

• PROPRIETARY DATA' MAY BE DISCLOSED TO THE ELECTRIC POWER RESEARCH INSTITUTE (EPRI), ELECTRIC UTluTY REPRE61NTATNES AND THEIR DIRECT CONSUL 1 ANTS, EXCLUDING DIRECT COMMERCIAL COMPETITORS, AND THE DOE NATONAL LABORATORIES UNDER THE PROHIBf70NS AND RESTRICTONS ABOVE.

(B) THi$ NOTICE SHALL BE MARKED ON ANY nEPRODUCTON OF THESE DATA, IN WHOLE OR N PART, O WESTINGHOUSE CLASS 3 (NON PROPRIETARY)

EPRI CONFIDENTIAL / OBLIGATION NOTICES:

NOTICE: 10 20 3 04 O5 O CATEGORY: A@ B OC DoOE O F- O

[3 DOE CONTRACT DELIVERABLES (DELIVERED DATA)

SUBJECT TO SPECIFIED EXCEPTONS, DtSCLOSURE OF THIS DATA 13 RESTRICTED UNTL SEPTEM0ER 30,1995 OR DESIGN CtRTIFICATION UNDER DOE CONTRACT DE ACO3-90SF18495, WHICHEVER IS LATER.

ORIGIN ATOR . ElGNATURE/DATE p Diane Antolovich k j hg fy / g V AP600 RESPONSIBLE MANAGER Brian A. McIntyre y [ -

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Form 58202D (f42)

EPRI CONFIDENT 1AUTY / ODUGATION NOTICES NOTICE 1:

The sta in this docun.ent is subject to no contdenbahty obkgabons.

NOTICE 2:

The data in this dacument is propnetary and confdental to Westnghouse Ewetne Corporabon and'or its Contractors. It is forwarded to recipient under an otAgabon of Conf &nce ard Trust for limited purposes only. Any use, cbsclosu'e to unauthortred persons, or copying of this document or pa'is thereof is prohibited except as agreed to in n&ance by the Elecific Power Research Inc.tttute (EPRI) and Wesunghoue+ Electne Corporat on. Remoient of this data has a d#y to inquira of EPRI and/or Wesbnghouse as to the uses of the information contained herein that are permitted.

NOTICE 3:

The data in this document is proprietary and cor:fdental to Westnghouse Elect /ic Corporabon and'or its Cont

• actors. It is forwarded to recipient under an obligation of Contdence and Trust for use only in evaluation tasks specifically authonzed by the Electne Power Research inshtute (EPRI). Any use, declosure to unauthonzed persons, or copying this document or parts thereof la prohibited except as agreed to in a*ance by EPRI and Westnghouse Electne Corporation. Recipient of this data has a dJty to inQJire of EPRI and/or Washnghouse as to the uses of the informahon contained herein that are permitted This document and any copies or exc.orpts thereof that may have been grsnerated are to be retumed to Wesbnghouse, drectly or through EPRI, when reque.ted to do no.

NOTICE 4:

The data in this document is propnetary and confdental to Wesbnghouse Electnc Corporabun and/or its contractors. It is being revealed in contdence and trust onty to Employees of EPRI and to certain contractors of EPRI for lirnited ovaluabon tasks authortred by EPRI. Any use, d,sclosure to unauthortred persons, or copying of this document or parts thoroof is prohibited. This Document and any copies or excerpts thereof that may have been generated are to be retumed to Westnghouse, drectly or through EPRI, when requested to da no.

NOTICE 5:

The data in this document is propnetary and contdenbal b Weshnghouse Electnc Corporaton and/or its Contractors. Access to this data is given in Confdence and Trust only at Wesbnghouse facilites for hmited evaluabon tasks assigned by EPRt.

Any use, dsclosure to unauth'>rized persons, orcopying of this document or parts thereofis prohibhad Neither this document nor any excerpts therefrom are to be removed from Wesbnghouse facilites.

EPRI CONF 1DENTIALITY / OBUGATION CATEGORIES CATEGORY 'A' (See Delivered Data)

Consists of CONTRACTOR Foreground Data that is contained in an issued report.

CATEGORY *D' (See Delivered Data)

Consists of CONTRACTOR Foreground Data that is not conta.ned in an issued report, except for computer programs.

CATEGORY 'C' Consists of CONTRACTOR Background Data except for computer programs.

CATEGORY 'D" Consists of computer programs developed in the course of performing the Work.

CATEGORY *E" Consists of computer programs developed pnor to the Effectve Date or after the Effectrve Date but outside the scope of the Wort.

CATEGORY 'F' Consists of administratrve plans and administrattve reports.

DEFINITIONS DEUVERED DATA Consists of documents (e g. speoficabons, drawings, reports) wtiich are generated under the DOE contract DE ACO3-90SFtB495, i

0010.FRM

O rg.,m Simplified Passive Advanced Light Water Reactor Plant Program AP600 TIER 1 MATERIAL Plant Description Inspections, Tests, Analyses, and Acceptance Criteria O

Prepared for U.S. Department of Energy San Francisco Operations Office DE AC03-90SF18495 Revision 1 December 15,1992 O @westiez#o#se

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- Table of Contents -

Table of Contents . . . . .. . ..... . .. .. ... ..i Table of Contents Tables . . . . . ..... . ..... .. . . . iv Table of Contents - Figures . . . . . . ... . . .. .. vil Ust of Effective Sections . . . .. .. . . . .. . .. ..... Vill List of Aeronymns . . . . .. . .... ....... ......... ..x Definitions . . .. . ... . ... . . . . . . . . . . .J*

1.0 INTRODUCTION

. .. .. . . . ... .. . ... . 1.0-1 2.0 GENERAL PLANT DESCRll' HON . . . .. . . . . 2.0-1 3.0 SYSTEM BASED TIER I MATERIAL . .... ... ... . ..... 3.0 1 O

, 3.1.1 FUEL liANDUNG AND REFUEUNG SYSITM . . . ... ....... .. 3.1.1 1 3.1.2 REACTOR COOLANT SYSTEM . ... ..... . . ....... 3.1.2 1 3.1.3 REACTOR SYSTEM . ,.. .. .... . .. ... . . ..... 3.1.3 1 3.2.1 AUTOMATIC DEPRESSURIZATION SYSTEM . .. . ..... ...... ...... 3.2.1-1 3.2.2 CONTAINMENT SYSTEM . ..... .. . . .... ..... . . ..... 3.2.2-1 3.2.3 PASSIVE CONTAINMENT COOUNG SYSTEM , . . . . ... .. 3.2.3 1 3.2.4 PASSIVE CORE COOUNG SYSTEM . . . . . . .,.... . .. . .......... 3.2.4-1 3.2.4.1 Passive Residual lleat Removal lleat Excbangers . .... ... . ... . 3.2.4-1 3.2.4.2 Core Makeup Tanks . . . .. . .. . . .. .... .. 3.2.4 2 3.2.4.3 Accumulators . .. ..... . ......... . ... .. . . 3.2.4-3 3.2.4.4 in-Containment Refueling Water Storage Tank and Containment Recirculation . ..... 3.2.4-3 I

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4r (f Revision: 1 Effective: 12/15/92 3.2.4.5 pil Adjustment Tank .. .... .. . ... ......... .... . 3.2.4-4 3.2.5 STEAM GENERATOR SYSTEM . . . . . ... . ...... ... . .. 3.2.5-1 3.2.6 MAIN CONTROL ROOM EMERGENCY ll ABITAlllLITY SYSTEM . .. . . .. 3.2.6-1 3.3.1 COMPONENT COOLING WATER SYSTEM . . . . ... . . .. . .. 3.3.1-1 3.3.2 CllEMICAL AND VOLUME CONTROL SYSTEM . .. .... .. ... ..... 3.3.2-1 3.3.3 STANDBY DIESEL FUEL OIL SYSTEM . .. ..... .. . . . 3.3.3 1 3.3.4 FIRE PROTECTION SYSTEM . . . . .. . . ... .. .... . 3.3.4-1 3.3.5 MECllANICAL llANDLING SYSTEM . . .. ... 3.3.5-1 3.3.6 PRIMARY SAMPLING SYSTEM . .... . . . .. .... ... 3.3.6 1 3.3.7 NORMAL RESIDUAL IIEAT REMOVAL SYSTEM , . .. .. ... ....... 3.3.7 1

("N 3.3.8 SPENT FUEL PIT COOLING SYSTEM . .. .. . .. .. .. ... . 3.3.8-1 3.3.9 SERVICE WATER SYSTEM . .... . ... ... . . 3.3.9-1 3.3.10 CONTAINMENT llYDROGEN CONTROL SYSTEM , . . . .. ... 3.3.10-1 l

3.4.1 MAIN AND STARTUP FEEDWATER SYSTEM ,. .... ... . . .. . 3.4.1-1 3.4.2 MAIN STEAM SYSTEM ... .,.. . . .. .. ... . ..... .. 3.4.2-1 3.5.1 DIVERSE ACTUATION SYSTEM , . ,. .. . . ..... .. . 3.5.1 1 3.5. DATA DISPLAY AND PROCESSING SYSTEM ....... . .. . ... .. . 3.5.2-1 3.5.3 INCORE INSTRUMENTATION SYSTEM .... ... ....... .. 3.5.3-1 3.5.4 PLANT CONTROL SYSTEM . . . . .... . . .. .. .. ... 3.5.4-1 3.5.5 PROTECTION AND SAFLTY MONITORING SYSTEM . . . . .. ... 3.55-1 3.5.6 RADIAllON MONITORING SYSTEM . . . ... .. . ... ... ........ , 3.5.6-1 W Westinghouse II

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3.6.1 MAIN AC POWER SYSTEM ... ... .. . ...... . ...... 3.6.1 1 3.6.2 NON-CLASS 1E DC AND UPS SYSTEM ...... . . . . .. .... . 3.6.2 1 3.ti.3 PLANT IJGitT1NG SYSTEM . . . . . . . . . . . . . ....... .. 3.6.3-1 3.6.4 CLASS lE DC AND UPS SYSTEM . ... . . . . ... 3.6.4-1 3.6.5 ONSITE STANDBY POWER SYSTEM . . .. . . . . .. . 3.6.5 1 3.7.1 NUCLEAR ISLAND NONRADIOACTIVE VEN'llLATION SYSTEM . . . . . .... 3.7.1 1 3.7.2 CENTRAL CHILLED WATER SYSTEM .. ... ... .. . ... . .. 3.7.2-1 3.7.3 ANNEX / AUX BUILDING NONRADIOAC'llVE VENTILATION SYSTEM . . . . . . . . . . 3.7.3-1 3.7.4 DIESEL GENERATOR BUILDING VENTILATION SYSTEM . . ........... 3.7.4 1 4.0 NON SYSTEM BASED TIER 1 MATERIAL . . . . . . ..... ........ .. 4.0 1 4.1 IlUMAN FACTORS ENGINEERING ...... . . .... . ...... . . . .. 4.1-1 4.2 SAFETY-RELATED P! PING ,. ... .... ..... . . .... . . 4.2 1 4.3 NUCLEAR ISLAND BUILDINGS . . . .. ... . . . .... 4.3-1 4.4 INTERFACE. ........ ... ..... ........ . . .......... .. 4.4-1 5.0 StrE PARAMETERS . . . . .. .... . ....... .... ............ 5.0-1 i

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Tables 3.1.1 1 Fuel llandlii,g Sy>:em Impections. Tests. Analyses and Acceptance Criteria . . . 3.1.1-3 3.1.N Reactor Coolant Sy stem Inspections. Tests, Analysee and Acceptance Criteria . . .. 3.1.2-3 3.1.3-1 Reactor Sptem inspections. Tests, Ana. lyses and Acceptance Criteria ... . . . . 3.1.3 3 3.2.1 1 Automatic Depressuriution System inspections, Tests. Analyses and Acceptance Criteria . 3.2.1 2 3.2.2-1 Containment System inspections. Tests. Analyses and Acceptance Criteria ...... . 3.2.2 2 3.2.2 2 Principal Containment Penetratiens . . .... . . . 3.2.2-4 3.2.3-3 Passive Containment Cooling System Inspections, Tests, Analyses and Acceptance Criteria . 3.2.3 3 3.2.4 1.1 Passive Core Cooling System (PRHR lleat Exchangers)

Inspections, Tests, Analyses and Acceptance Criteria .... .. . . . 3.2.4-5 t

3.2.4 1.2 Passive Core Cooling System (Core Makeup Tanks)

Inspections, Tests. Analyses and Acceptance Criteria . . .. . . .. 3.2.4-8 3.2.4 1.3 Passive Core Cooling System (Accumulators)

Inspections. Tests. Analyses and Acceptance Criteria . . . . . ... .. 3.2.4-13 3.2.4-1.4 Passive Core Cooling System (IRWST and Containment Recirculation)

Inspections, Tests, Analyses and Acceptance Criteria . . . . . . .. 3.2.4 15 3.2.4-1.5 Passive Core Cooling System (pH Adjustment Tank)

Inspections, Tests, Analyses and Acceptance Criteria . . .. . .. . . . 3.2.4-19 3.2.51 Steam Generator System inspections. Tests, Analyses and Acceptance Criteria . . .. 3.2.5-3 1

3.2.6-1 Main Control Room Emergency Habitability System inspections. Tests, l Analyses and Acceptance Criteria . .. .. . . . . ,. 3.2.6-2 l

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h Revision:1 Effective: 12/15/92 3.3.1 1 Component Cooling Water System inspections, Tests, Analyses ar.d Acceptance Criteria , 3.3.1-2 i

3.3.21 Chemical and Volume Control System inspecticns. Tests, Analyses and Acceptance Criteria 3.3.2 3 l l

3.3.31 Standby Diesel Fut Oil System inspections. Tests, Analyses and Acceptance Criteria . 3.3.3-2 3.3.4-1 Fue Protection System inspections, Tests Analyscs and Acceptance Criteria . , , 3.3.4-2 3.3.51 hicchanical llandling System Inspections, Tests, Analyses and Acceptance Criteria . 3.3.5-2 l 3.3.6-1 Primary Sampling System inspections, Tests, Analyses and Acceptance Critena . . 3.3.t;-2 3.3.71 Normal Residual lleat Removal System inspections, Tests. Analyses and Acceptance Criteria 3.3.7 3 3.3.81 Spent Fuel Pit Cooling System Inspections, Tests, Analyses and Acceptance Criteria . . . 3.3.8-2 3.3.91 Service Water System inspections, Tests, Analyses and Acceptance Criteria . . . 3.3.9-2 3.3.10-1 ll>drogen Control System Inspections, Tests, Analyses and Acceptance Criteria . . . . 3.3.10-2 3.4.1-2 3 3.4.1-1 hiain and Stanup Feedwater System Inspections, Tests, Analyses and Acceptance Criteria

[V 3.4.2-1 hiain Steam Systeu Inspections, Tests. Analyses and Acceptance Criteria .. . .. 3.4.2-2 3.5.1-1 Diverse Actuation System inspections. Tests. Analyses and Acceptance Criteria . ..... 3.5.1 2 3.5.2-1 Data Display and Processing System Inspections, Tests, Analyses and Acceptance Criteria . . 3.5.2-2 3.53-1 Incore Instrumentation System inspections, Tests, Analyses and Acceptance Criteria . .. '.s.5.3-2 3.5.41 Plant Control System Inspections, Tests, Analyses and Acceptance Criteria . . .... 3.5,4-2 3.5.51 Protection and Safety hionitoring System Inspections, Tests, Analyses and Acceptance Criteria 3.5.5-4 3.5.61 Radiation hionitoring Sy stem inspections Tests, Analyses and Acceptance Criteria . 3.5.6-2 3.6.1 1 hiain AC Power System inspections, Tests, Analyses and Acceptance Criteria . .. . 3.6.1-2 3.6.2-1 Non. Class IE DC and UPS System Inspections, Tests, Analyses and Acceptance Criteria , 3.6.2-2 3.6.3-1 Plant Lighting System inspections, Tests, Analyses and Acceptance Criteria , . . 3.6.3-2 s

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(,) Revision: 1 Effective: 12/15/92 e 3.6.4-1 Class IE DC and UPS System inspections, Tests. Analyses and Acceptance Criteria .. . 3.6.4-2 3.6.5-1 Onsite Standby Power System inspections, Tests. Analyses and acceptance Criteria . ... 3.6.5-2 3.7.1-1 Nuclear Island Nonradioactive Ventilation System Inspections, Tests.

Analyses and Acceptance Criteria .... . . . . . . .... 3.7.1-2 3.7.21 Central Chilled Water System Inspections, Tests, Analyses and Acceptance Criteria .. . 3.7.2-2 3.7.3-1 Annex / Aux Buildings Nonradioactive Ventilation System . . .. . . .. . 3.7.3-2 3.7.4-1 Diesel Generator Building Ventilation Systern Inspections, Tests.

Analyses and Acceptance Criteria .. .. . . .. . . 3.7.4 ~.

4.1 Human Factors Engineering laspections, T-tt, Analyses, and Acceptance Criteria . 4.1-2 4.2 Safety-related Piping Inspections, Tests, Anaiyses, and Acceptance Criteria . . . . 4.2-2 4.3 Nuclear Island Buildingt .. . ... . .. . .. . 4.3-3 5.0-1

,A) i 5.0 Site Parameters . ... ,. . . . .

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Figures 5.0-1 Safe Shutdown Earthquake, Horizontal Design Response Spectra . ... 5.0-2 5.0-1 Safe Shutdown Earthquake, Vertical Design Response Spectra . . . . 5.0-3 NSa; I

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e List of Acronyms ac - Alternating current AHU - Air handling unit 11tu - liritish thermal unit Class IE - Safety Class IE CMT - Core makeup tank dc - Direct current eff Effective gpm - Gallons per minute HEPA - High efficiency particulate air HVAC - Heating, ventilation and air conditioning Hz - Hertz IRC - Inside reactor containment IRWST - In-containment refueling water stonge tank ITAAC - Inspections, Tests, Analyses and Acceptance Criteria kva - Kilo volt amperes LOCA - Loss of coolant accident t AISIV - hfain steam isolation valve 1

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N Symbol Stamp NPSH

- ASNIE Code symbol stamp

- Net positive suction head NIT Symbol Stamp - AShtE Code symbol stamp ORC - Outside reactor containment pH - Hydrogen ion concentration (negative logarithm)

PORY - Power operated relief valve PRHR - Passive residual heat removal psid - Pounds per square inch differential psig - Pounds per Square inch gravity PZR - Pressurizer RHR - Residual heat removal RCP Reactor coolant pump sefm - Standard cubic feet per minute TBD - To be determined UPS - Uninterruptable power supply Il W

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Effective: 12/15/92 i Definitions Defense-in-depth The defense-in depth systems are those nonsafety-related active systems that:

1. Directly act to prevent unnecessary actuation of the safety-related passive systems EXCEFf:

a. Where a specific defense-indepth function provides an insignificant or limited benefit, or

b. Where actuation of the passive safety-related system is not onerous (such as actuation of the passive containment c,oling system)

2. Provide support functions (such as heat removal or electrical power) to the nonsafety-related systems captured by the criteria above.

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U Safets-relate 4 A classification applied to items relied upon to remain functiceal during or following a design basis event to provide a safety-related function. Safety-related also applies to documentation and services affecting a safety-related item.

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. t List of Effective Sections SECTION TITLE REVISION DATE 3.1.1 FUEL HANDLING AND REFUELING SYSTEM Rev. 01 12/15/92 3.1.2 REACTOR COOLANT SYSTEM Rev. 01 12/15/92 3.1.3 REACTOR SYSTEM Rev. 01 12/15/92 3.2.1 AUTOMATIC DEPRESSURIZATION SYSTEM Rev. 01 12/15/92 3.2.2 CONTAINMENT SYSTEM Rev. 01 12/15/92 3.2.3 PASSIVE CONTAINMENT COOLING SYSTEM Rev. 01 12/15/92 3.2.4 PASSIVE CORE COOLING SYSTEM Rev. 01 12/15/92 3.2.5 STEAM GENERATOR SYSTEM Rev. 01 12/15/92 3.2.6 MAIN CONTROL ROOM EMERGENCY H ABITABILITY SYSTEM Rev. 01 12/15/92

'% 3.3.1 COMPONENT COOLING WATER SYSTEM Rev. 01 12/15/92 3.3.2 CHEMICAL AND VOLUME CONTROL SYSTEM Rev. 01 12/15/92 3.3.3 STANDBY DIESEL FUEL OIL SYSTEM Rev. 01 12/15/92 3.3.4 FIRE PROTECTION SYSTEM Rev. 01 12/15/92 3.3.5 MECH ANICAL llANDLING SYSTEM Rev. 01 12/15/92 3.3.6 PRIMARY SAMPLL .G SYSTEM Rev. 01 12/15/92 3.3.7 NORM AL RESIDUAL HEAT RFMOVAL SYSTEM Rev. 01 12/15/92 3.3.8 SPENT FUEL PIT COOLING SYSTEM Rev.01 12/15/92 3.3.9 SERVICE WATER SYSTEM Rev. 01 12/15/92 3.3.10 CONTAINMENT HYDROGEN CONTROL SYSTEM Rev. 01 12!15/92 3.4.1 MAIN AND STARTUP FEEDWATER SYSTEM Rev. 01 12/15!92 3.4.2 MAIN STEAM SYSTEM Rev. 01 12/15/92 f) W Westinghouse

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Effective: 12/15/92 SECTION TITLE REVISION DATE 3.5.1 DIVERSE ACTUATION SYSTEM Rev. 01 12/15/92 3.5.2 DATA DISPLAY AND PROCESSING SYSTEM Rev. 01 12/15/92 3.5.3 1NCORE INSTRUMENTATION SYSTEM Rev. 01 12/15/92 3.5.4 PLANT CONTROL SYSTEM Rev. 01 12/15/92 3.5.5 PROTECTION AND SAFETY MONITORING SYSTEM Rev. 01 12/15/92 3.5.6 RADIATION MONITORING SYSTEM Rev. 01 12/15/92 3.6.1 MAIN AC POWER SYSTEM Rev. 01 12/15/92 3.6.2 NON-CLASS 1E DC AND UPS SYSTEM Rev. 01 12/15/92 3.63 PLANT LIGHTING SYSTEM Rev. 01 12/15/92 3.6.4 CLASS 1E DC AND UPS SYSTEM Rev. 01 12/15/92 3.6.5 ONSITE STANDBY POWER SYSTEM Rev. 01 12/15!92 3.7.1 NUCLEAR ISLAND NONRADIOACTIVE VENTILATION SYSTEM Rev. 01 12/15/92 3.7.2 CENTRAL CillLLED WATER SYSTEM Rev. 01 12/15/92 3.7.3 ANNEX / AUX BUILDING NONRADIOACTIVE VENTILATION SYSTEM Rev. 01 12/15/92 3.7.4 DIESEL GENERATOR BUILDING VENTILATION SYSTEM Rev.01 12/15/92 W Westinghouse

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1.0 INTRODUCTION

The Tier 1 Design Certification material in this -

document is the AP600 design certified by the U.S.

Nuclear Regulatory Commission, published by reference in an appendix to 10 CFR Part $2. It includes a

' description of the principal design basis and principal

-features of the certified design and the inspections, tests, analyses and acceptance criteria which are necessary and sufficient to provide reasonable assurance that, if the inspections, tests and analyses are performed and the

- acceptance criteria are met, a plant which references the design is built and will operate in accordance with the-design certification.

~ 'Ihis Tier 1 Design Certification material - is

-presented for each of the system-based safety related and defense-in-depth functions.-

This Tier 1 Design Certification material includes the interface requirements to be met by those portions of the plant for which the application does not seek design certification, together with justification that compliance with' the interface requirements is verifiable through

- inspections, tests or analyses.

The Design Certi0 cation material also includes the site parameters postulated for the oesign.

'A larger body of material submitted as part of the'

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-AP600 design certificaiion application is documented in '

the AP600 Standard Safety Analysis Report and is considered Tier 2 unterial. ' This Tier 2 material is not included in this document. ,

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( 4 GENERAL PLANT DESCRIPTION Revision: 1 NE E"

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Effective: 12/15/92 2.0 GENERAL PLANT DESCRIPTION The AP600 is a Westinghouse pressurized water reactor The reactor coolant system is located within the auclear power plant which produces nominal 600 containment building and consists of two main cculant rr.egawatts net electrical output. The plant incorporates loops, a reactor vessel, two steam generators, four passive safety systems which provide essential reactor canned motor reactor coolant pumps and a pressurizer.

protection functions. The steam and feedwater lines are routed from the steam generators to containment penetrations below the PRINCIPAL DESIGN PARAMETERS operating deck. They penetrate the containment vessel and are routed through the main steam isolation valve Nuclear Steam Supply System area in the auxiliary building to the turbine building.

Power Rating 1940 M Wt ne passive core cooling system is also located in Reactor Core Power Rating 1933 M Wt the containment building.

Reactor Coolant Pressure Boundary A fuel transfer system is provided to transfer Design Pressure 2485 psig nuclear fuel assemblies between the refueling cavity in Reactor Coolant Pressure Boundary the containment building and the fuel transfer Design Temperature 650' F canal / spent fuel pit located in the fuel handling area of Containment Internal Design Pressure 45 psig the auxiliary building.

Access to the containment ! provided through two personnel airlocks, the main equipnwnt hatch and a

[V) PLANT ARRANGEMENT AND STRUCTURES maintenance hatch.

The shield building serves as radiation shielding and The plant is comprised of six principal building a missile barrier and is a part of the passive containment structures: cooling system. It also is a major structural member for the nuclear island. Floor slabs and structural walls of

• Nuclear island the auxiliary building are structurally connected to the

• Turbine building cylindrical section of the shield building.

• Annex 1 building The reinforced concrete shield building roof

. Annex 11 buildig supports the passive containment cooling system water

• Diesel generator building storage tank and air diffuser. Air intakes are located

• Radwaste building. along the cylindrical portion of the shield building.

The auxiliary building houses the safety-related The nuclear island consists of a containment mechanical and electrical equipment located outside the building, a shield building and an auxiliary building. containment and shield building. It also provides The foundatior, for the nuclear island is an integral radiation shielding for the radioactive equipment and basemat which supports these buildings. piping that is housed within the building.

The containment building is the containment vessel The auxiliary building includes the following:

and the structures contained within the containment vessel. He containment vessel is a freestanding

• Main control room cylindrical steel containment vessel with elliptical upper

• Instrumentation and control equipment area and lower heads. It is surrounded by a reinforced a Class IE de & UPS equipment area concrete shield building.

• Fuel handling area

,3 C W Westinghouse

Ti:t 1 D: sign CIrtific: tion M:t: rial GENERAL PLANT DESCRIPTION .

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Revision: 1 Effective: 12/15/92

• Mechanical equipment areas

• Containtnent penetration areas

• Main steam and feedwater isolation valve compartment.

The annex ! and II buildings provide access for personnel and equipment to the clean areas of the _

auxiliary building and to the radiologically controlled area. The buildingsinclude bcalth physics facilities and provides permnnel and equipment access to and from the containment building and the rest of the radiologically controlled area via the auxiliary building.

The annex 1 and II buildings also contain the non-Class IE ac and de electric power systems, other electrical equipment, the technical support center and various 1 heating, ventilation and air conditioning systems, The diesel generator building houses two diesel generators. These generators provide backup power in case of disruption of normal ac power sources.

1he turbine building houses the main turbtne, -

generator and associated fluid and electrical systems and various plant support systenu.

The radwaste building houses equipment associated with the collection and processing of solid radioactive waste generated by the plant.

t 2.0-2 W Westingflouse l

Ti;r 1 D: sign C rtific;ti:n M;t:rl:1 c,

( ) SYSTEM TIER 1 MATERIAL *'

• O' Revision: 1 Effective: 12/15/92 3.0 SYSTEM TIER 1 MATERIAL

'This section provides Tier 1 material for each of the system-based safety-related and defense-in-depth functions. The systems for which Tier i material is provided, have been grouped into subsections. Each subsection contains systems which have similar or related plant functions. The systems within each subsection are presented in alphabetical order by system acronym. Each system has c Tier 1 Design Description and the associated Inspections. Tests, Analyses and Acceptance Criteria (ITAAC).

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Ti r 1 D:sion Certification Mit:rint n

i I FUEL HANDLING AND REFUELING SYSTEM s

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J Effective: 12/15/92 3.1.1 FUEL HANDLING AND REFUELING SYSTEM Design Description The fuel handling and refueling system (FHS) core components and other equipment moved as part of handles and stores new and spent fuel. The fuel the refueling process, handling and refueling system is a nonsafety-related system except for the following safety-related functions New Fuel Storage outside containment:

New fuel is stored in a high density rack which

• Prevent dropping or jamming of fuel assemblies includes integral neutron absorbing material to maintain during transfer operation, the required degree of subcriticality. De rack is designed to store fuel of the maximum design basis

• Prevent dropping of fuel handling devices during the enrichment. The rack is located in the new fuel storage fuel transfer operation, pit. He pit is located in the fuel handling area of the auxiliary building.

• Preserve the integrity of the containment pressure The rack in the new fuel pit consists of an array of boundary where the fuel tnmsfer system penetrates cells interconnected to each other at several elevations the containment. The containment isolation safety- and to supporting grid structures at the top and bottom.

related function is covered in the containment Spacing between the cells provides a separation between

' system Tier 1 Design Description in Section 3.2.2. adjacent fuel assemblies and with the absorber raterial is sufficient to maintain a suberitical configuration.

• Prevent criticality during fuel handling operations by nmintaining the geometrically safe configuration Spent Fuel Storage of the fuel Spent fuel is stored in high density racks which

• Store new fuel in a manner to maintain the required include integral neutron absorbing material to maintain degree of subcriticality the required degree of subenticality. He racks are designed to store fuel of the maximum design basis

• Store spent fuel in a manner to maintain the enrichment. The spent fuel racks are located in the required degree of suberiticality spent fuel pool. The spent fuel pool is located in the auxiliary building fuel handling area.

The fuel handling and refueling system performs the Each rack in the spent fuel pool consists of an array following defense-in-depth function: of cells interconnected to each other at several elevations and to supporting grid structures at the top and bottom.

• Limit spent fuel lift height so that the minimum A gated opening connects the spent fuel pool and required depth of water shielding is maintained, fuel transfer canal. The fuel transfer canal is connected to the in-containment refueling cavity by a fuel transfer The system includes handling equipment and tools tube. The spent fuel transfer operation is completed and storage racks for both new and spent fuel. The underwater with a level of shielding water maintained system also includes equipment for handling and storing above the spent fuel assemblies. The gate assembly that separates the spent fuel pool and fuel transfer canal

/m i W Westin=e, house 3.1.1-1 m

Tier 1 DIsign C;rtific: tion M"talal FUEL HANDLING AND REFUELING SYSTEM iiih Hilj Revision: 1 T Effective: 12/15/92 allows the fuel transfer canal to be drained without reducing the water level in the spent fuel pool.

Fuel Handling The fuel handling equipment handles the spent fuel assemblies underwater from the time they leave the reactor sessel until they are placed in a container for shipment from the site. The handling equipment used to raise and lower spent fuel has a limited maximum lift height so that a minimum depth c,4 shielding water is maintained. The fuel handling devices have provisions to avoid dropping orjamming of fuel assemblies during transfer operation. The handling equipment has provisions to avoid dropping of fuel handling devices during the fuel transfer operation.

The fuel transfer system is used to move fuel assemblies between the containment building and the auxiliary building fuel handling area. In the fuel handling area. fuel assemblies are moved by the fuel handling machine and spent fuel handling tool.

The fuel handling nachine and spent fuel handling tool in the auxiliary building fuel handling area are safety-related components. The refueling machine in the containment is not safety-related.

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sm m U U Tier 1 Design Certification Material U

FUEL HANDLING AND REFUELING SYSTEM g- .

Revision: 1 Effective: 12/15/92 7 Table 3.1.1 Fuel Handling System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptar'ce Criteria

1. He required degree of subcriticality is Using as-built dimensions and material, a ne required degree of suberiticality is maintained in new fuel storage racks. criticality analysis shall be performed. maintained. K,., is less than or equal to 0.95 for dry storage with new fuel of the maximum enrichment. K ais less than or equal to 0.98 for the postulated accident of flooding the new fuel storage area with unborated water.

2. The required degree of suberiticality is Using as-built dimensions and material, a De required degree of suberiticality is maintained in spent fuel storage racks. criticality analysis shall be performed. maintained. K,is less than or equal to 0.95 with unborated water for conditions including postulated fuel handling accidents.

3. He I:ft controls for the handling equipment Functional testing of the fuel handling equipment Prevents the top of the fuel assembly from used to raise and lower spent fuel limits lift shall demonstrate that minimum water shielding rising within [TBD] feet of the water surface.

height so that the minimum required depth can be maintained.

of shielding water is maintained.

4. He fuel handling machine and spent fuel a. Using a dummy fuel asseinbly the fuel a. Fuel handling devices grip and do not handling tool and the fuel transfer system handling interlocks shall be functionally release under load.

have prosisions to prevent dropping or tested to show that a fuel assembly does not jamming of fuel assemblies during transfer drop.

P#'" "' b. Using a dummy fuel assembly, the fuel b. The fuel transfer system does not jam.

transfer system shall be functionally tested to show that it does not jam.

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Tier 1 Design Certification Material FUEL HANDLING AND REFUELING SYSTEM Revision: 1 Effective: 12/15/92 _

Table 3.1.1 Fuei Handling System Inspections, Tests, Analyses and Accertance Cnteria 1

Certified Design Commitment e Inspections Tests, Analyses Acceptance Criteria

5. The fuel handling machine and spent fuel Load testing of the handling equipment shall be A test load is lifted by the handling equipment.

handling tool include provisions to prevent perform-d at 1.25 times the rated load to show dropping of fuel handling devices during the that a fuel handling device will not drop.

fuel transfer operation.

W Westinghouse

TI:t 1 D:si:n Ccrtific: tion M:t: rial FIEACTOR COOLANT SYSTEM "~ ,

(V) Revision: 1  %

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E f fective: 12/15/92 3.1.2 REACTOR COOLANT SYSTEM Design Description 1he reactor coolant xystem (RCS)is a safety-related

• The pressurizer maintains the system pressure system and is a barrier designed to prevent the release during operation. During the reduction or increase of Gssion pmducts to the atnuphere. The teactor of plant load, the pressurizer accommodates volume cwlant system provides the following safety-related changes in the reactor coolant and limits pressure functions: transients

• 1he reactor coolant system including the major The reactor coolant system consists of two heat components, steam generator channel head, tubes, iransfer circuits, each with a steam generator, two and tubesheet and interconnecting piping provides reactor coolant pumps, a single hot leg and two cold the pressure boundary for the reactor coolant. legs for circulating reactor coolant and the supports for the steam generator and reactor vessel. In addition, the

• The shell side of the steam generator provides the system includes the pressurizer, interconnecting piping, pressure boundary for the secondary fluid and valves, and instrumentation for operational control and

,. 3 connects to the steam generator system. Refer to safeguards actuation.

the steam generator system Tier 1 Design During operation, the reactor coolant pumps

[V} Description in Section 3.2.5, circulate pressurized water through the reactor vessel and the steam generators. The water, which serves as

• The reactor coolant system in conjunction with the coolant, ruoderator, and solvent for boric acid, is heated passive core cooling system provides circulation of as it passes through the core and transfers heat to the coolant when the reactor coolant pumps do not steam generator system. The water then is retumed to operate. 'he reactor vessel by the pumps to repeat the process.

The reactor coolant system pressure is maintained

• The reactor coolant system provides now coast by the pressurizer and by the use of electrical heaters or down on loss of power to the reactor coolant pressuriter spray, pumps. Safety valves are connected to the pressurizer to provide overpressure protection for the reactor coolant

• The reactor coolant system includes overpressure system.

protection. The valves that isolate auxiliary systems that support the reactor coolant system are part of the reactor coolant The reactor coolarit mystem provides the following pressure boundary.

defense-in-depth func ans: The reactor coolant system includes the following major compotients:

• The reactor coolant system provides circulation of reactor coolant through the reactor vessel and the

• The reactor coolant pumps. Four canned motor primary side of the steam generator transferring pumps pump Huid through the reactor coolant and heat to the secondary side of the steam generator.

3.1.2 1 bT J W-Westinghouse

Ti:r 1 Dssign Certification Mst rial REACTOR COOLANT SYSTEM

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Revision: 1 Effective: 12/15/92 reactor systems. Two pumps are coupled with each The pressurizer is used for reactor coolant system steam generator, pressure control by maintaining liquid and vapor at saturated conditions. A spray nozzle and two nozzles

• Two steam generators. for the safety and depressurization valve inlet headers are located in the top head of the pressurizer. Electrical

• De pressurirer which is attached by the surge line heaters are installed in the lower portion of the to one of the reactor coolant hot legs. He spray pressurizer, He bottom head contains the surge line lines connect the pressurizer and cold legs. norr.le. The surge line connects the pressurizer to a hot leg, and provides for the flow of reactor coolant into

• The safety valves and automatic depressurization and out of the pressurizer during reactor coolant system system valves. Refer to the automatic thermal expansions and contractions, depressurization system Tier 1 Design Description ne piping, fittings, and valves connecting auxiliary in Section 3.2.1. or support systems are part of the reactor coolant pressure boundary. Pressure boundary integrity

• The reactor vessel head vent isolation valves. requirements for these items out to the end of the reactor coolant pressure boundary conform to the

• The interconnecting piping and fittings between the requirements for the reactor coolant system.

preceding principal components. He reactor vessel supports are located beneath the inlet nozzles. He vessel supports transfer loads from he reactor vessel and control rod drive mechanism the reactor pressure vessel to the containment interior housings which are part of the reactor coolant pressure structure.

boundary are part of the scactor system. Refer to the ne steam generator support includes a vertical reactor system Tier 1 Design Description in column extending from the steam generator Sectica 3.1.3. compartment floor to the bottom of each steam The steam generator is a vertical shell and U-tube generator channel head, evaporator with integral moisture separating equipment. He lower horizontal steam generator support is The steam generator tubes are welded to the tubesheet located at the channel head and permits movement of the cladding. Tubes are accessible for remote inspection or steam generator during plant heatup and cooldown. The other maintenance activities. upper horizontal steam generator support is located on The reactor coolant pumps are high-inertia canned the lower shell below the transition cone. The generator motor pumps that circulate the reactor coolant through loads are transferred to the compartment wall.

the reactor vessel, loop piping, and steam generators. Le pressurizer is supported by the connection of Reactor coolant system piping is configured with the integral support skirt into the supporting structure.

two main coobnt ' oops, each of which employs a single A support near the top of the pressurizer provides lateral hot leg pipe to transport reactor coolant to a steam support and transfer loads to the containment interior generator. The two reactor coolant pump suction structure.

nozzles are welded to the outlet nozzles on the bottom of the steam generator channel head. Two cold leg pipes in each loop (one per pump) transport reactor coolant back to the reactor vessel to complete the circuit.

3.1.2-2 3 Westinghouse .

(3 v (3 v (v~b Tier 1 Design Certification Material REACTOR COOLANT SYSTEM -

2 Revision: 1 Effective: 12/15/92 _ .. .

si Table 3.1.2 Reactor Coolant System i Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests Analyses Acceptance Criteria

1. He reactor cochnt system pres <ure he reactor coolant pressure boundary shall be he reactor coolant system has an N Symbol boundary contains the reactor coolant under hydrostatically tested. Stamp and an NPT Sy mbol Stamp.

operating pressures and temperatures.

2. He steam generator secondary side contains ne steam generator secondary boundary shall Re steam generator has an N Symbol Stamp.

secondary fluid under operating pressures be hydrostatically tested.

and temperatures.

3. He reactor coolant sy stem circulates coolant During hot functional testing a flow test of the ne total measured cold leg Cow rate during at a flow rate consistent with component reactor coolant system Cow rate shall be the hot functional tests with reactor coolant designs. performed at a pressure of [TBD] and pressure between [TBD] and temperature temperature of [TBD}. between (TBD] is a minimum of ITBD}.

4. Safety valves provide overpressure a. Inspections shall be conducted to confirm a. The sum of the rated capacities of the safety protection for the reactor coolant sy stem _ that the value of the vendor code plate rating valves connected to the pressurizer exceeds is greater than or equal to system relief [TBD] Ibrhr as recorded on the valve vendor requirements. code plates.

b. De safety valves shall be tested to b. He set pressure of the vahes is :

demonstrate that the valves open at the set pressure. [TBD]psig [TBD] psig W-Westincrhouse a

A O Tier 1 Design Certification Material REACTOR COOLANT SYSTEM $ -

Revision: 1 Effective: 12/15/92 _

i Table 3.1.2 Reactor Coolant System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

5. De steam generators are supported by The supports shall be inspected to show that the He steem generator is supported by a column column below the channel head, a lower locations are correct. below the steam generator and the two horizontal support located at the channel horizontal supports.

head and the upper horizontal support located on the lower shell below the tritn3ition core.

6. He pressurizer is supported by an integral The supports shall be inspected to show that ne bolts in the support skirt are installed and support skirt connected to the supporting they are connected to the support structure. the upper support is installed.

structure and a support near the top of the pressurizer.

7. De reactor coolant system provides flow Using as-built information the pump inertia shall Re pump inertia shall be greater than or equal coast down on loss of power to the reactor be determined. to iTBD] foot-pounds.

coolant pumps.

8. The pressurizer maintains reactor coolant a. During hot functional testing with the water a. He pressurizer temperature rises at a rate system pressure. level in the pressurizer between [TBD] and greater than [TBD]*/hr.

[TBD] and the water temperature between

[TBD] and [TBD], the pressurizer heaters shall be energized.

b. During hot functional testing, with the b. He pressurizer pressure shall be reduced at pressure between [TBD] and [TBD] and the a rate greater than [TBD] psig/ min.

pressurizer temperature between [TBD] and

[TBD] the pressurizer spray shall be actuated.

W Westinghouse

Ti:r 1 D: sign Certific: tion M;t; rial REACTOR SYSTEM (V) Revision: 1 1 '..

4" Effective: 12/15/92

- bd 3.1.3 REACTOR SYSTEM Design Description The reactor system (RXS) is a safety-related system fuel assemblies are also part of the reactor system. The and is a barrier designed to prevent the release of fission reactor system maintains a cootable geometry during products to the atmosphere. normal operation, anticipated transients; and design he reactor system includes the reactor pressure basis accident conditions.

vessel, fuel, reactor internals, reactor control The reactor system includes the following major assemblies, the control rod drive mechanisms, the components:

conduit for the incore instrumentation, and miscellaneous other equipment within or attached to the Reactor vessel reactor pressure vessel incleaing those components that are in the integrated head package. He reactor vessel consists of a cylindrical section ne reactor system provides the following with a hemispherical bottom head and a removable safety-related functions: flanged hemispherical upper head. %e upper head is attached to the vessel by studs. He vessel has four

• The reactor system prosides a pressure boundary to inlet nonles, two outlet nonles, and two direct contain the reactor coolant. injection mules.

He reactor vesul is fabricated from low .dloy steel (V)

• De reactor vessel provides support and alignment and clad with corrosion resistant material Ferritic for the reactor internals and core, reactor vessel materials comply with the fracture toughness requirements of Section 50.55a and

• The reactor vessel, the reactor internals, and .sel Appendices G and H of 10 CFR 50.

assemblies direct main coolant flow through the The primary inlet and outlet noules are connected core. to the reactor coolant loop piping. The direct vessel injection nonles are connected to the direct vessel

• The reactor vessel provides support end alignment injection line. Rese nonles are located in the reactor for the control rod drive mechanisms. vessel upper section. He inlet and outlet inonles are offset at different elevations There are no penetrations

• Re reactor vessel provides support and alignment in the reactor vessel below the elevation of the core.

for the in. core instrumentation assemblies. The reactor vessel supports the internals. The interfaces between the reactor vessel and the lower

• Electrical power interruption initiates release of the internals core barrel are such that the main coolant flow control rod drive assembly. enters through the inlet nonle md is directed down through the annulus between the reactor vessel and core

%e reactor system includes those components that barrel and flows up through the core.

contain the reactor coolant around the core, direct Penetrations in the upper head provide access for the coolant flow through the core, and control the rate of control rod drive mechanisms. Penetrations for the the nuclear reaction in the core. The reactor system incore instrumentation and head vent are also in the receives water from and delivers, at increased upper head.

temperature, water to the reactor coolant system. The O 3.1.3-1 (v) W Westinghouse

Tizr 1 Design Certific: tion Msttrial REACTOR SYSTEM n!! Revision: 1 Effective: 12/15/92 e

Reactor internals Control Rod Drive Mechanisms

%e reactor internals consist of two assemblies - the Control rod drive mechanisms are located on the lower internals and the upper internals. ne reactor head of the reactor vessel. Le primary functions of the internals provide the protection, alignment and support control rod drive mechanism is to insert or withdraw, for the core, and neutron absorbing rods to provide safe rod cluster control assemblies and gray rod cluster and reliable reactor operation. In addition, the reactor assemblies from the core to control reactivity, internals help direct the main coolant flow to and from ne control rod drive mechanism consists of four the fuel assemblies; absorb control rod dynamic loads, separate subassemblies. These are the pressure vessel, fuel assembly loads, and other loads and transmit these coil stack assembly, latch assembly, and drive rod loads to the reactor sessel; and support instrumentation assembly. The pressure vessel includes a latch housing within the reactor vessel, and a rod tras sl housing.

The major support member of the reactor internals The control rod drive mechanism releases the is the lower core support assembly. This assembly driveline if electrical power is interrupted. When includes the core barrel, lower core support plate, released from the control rod drive mechanism, the secondary core support, radial reflectors, radial drive line falls by gravity into a fully inserted position.

supports, and related attachment hardware. The lower core support assembly is supported at its upper flange Integrated head package from a ledge in the reactor vessel flange. The restraint system for the lower end of the core barrel restricts The integrated head package includes conduit for the rotational and translational movement, but allows for incore instrumentation, cables and connectors for the radial thermal growth and axial displacement, incore instrun entation system, seismic restraint for The neutron reflector assemblies are located inside reactor head vent piping and valves and the incore the core barrel and above the lower core support. instrumentation conduit, seismic restraints for control These assemblies form the radial periphery of the core, rod drive mechanisms, and supports for cables including Loads acting vertically downward and transverse the messenger tray and cable bridge, loads are carried by the lower core support assembly to the reactor vessel.

He upper core support assembly consists of the upper support, the upper core plate, the support columns, and the guide tube assemblies.

The support columns establish the spacing between the upper support and the upper core plate. He support columns transmit the mechanical loadings between the two plates. He support cohimns provide a protective path for the incore detectors during installation and reactor operation.

The guide tube assemblies sheath and guide the control rod drive shafts and control rods.

3.1,3-2 W-Westinohouse a

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J G Tier 1 Design Certification Material REACTOR SYSTEM Revision: 1 Effective: 12/15/92 _

Table 3.1.3 Reactor System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment Inspections, Tests, Analyses Acceptance Criteria

1. He reactor vessel provides a pressure ne reactor vessel shall be hydrostatically He vessel has an N Symbol Stamp.

boundary to contain the reactor coolant. tested.

2. The reactor sessel directs main coolant flow During hot functional testing a flow test of the ne total measured cold leg flow rate during 4 through the core by close interface with the reactor coolant system flow rate shall be per- the hot functional tests with reactor coolant reactor internals. formed at a pressure of [TBD] and temperature pressure between [TBD] and temperature of [TBD}. between [TBD) is a minimum of [TBD].

3. He pressure housing for the control rod The control rod drive mechanism pressure hous- De control rod drive mechanism has an NI'T drive mechanism provides a pressure ing shall be hydrostatically tested. Symbol Stamp.

boundary to contain the reactor coolant.

4. Electrical power interruption initiates release Stepping tests shall show movement of latches latches release when the respective coil is of the control rod drive assembly. for each coil. deenergized.

5. Feeitic reactor vessel materials comply with Tests of the reactor vessel material shall be The initial Charpy V-notch minimum upper the fracture toughness r quirements of performed to determine the upper shelf fracture shelf fracture energy level for the reactor vessel Section 50.55a and Appendices G and H of energy level. beltline material is at least 75 foot-pounds.

10 CI'R 50.

6. He reactor vessel is supported by supports De supports shall be inspected to show that the Re reactor vessel is supported on the inlet under the inlet noz2les. location is correct. nor21es.

3.1.3-3

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Ti:t 1 D: sign Ctrtificition Mitirl;l

.(O) . AUTOMATIC DEPRESSURIZATION SYSTEM ,

V Revision: 1 i H

Effective: 12/15/92 3.2.1 AUTOMATIC DEPRESSURIZATION SYSTEM Design Description The automatic depressurization system (ADS) is a

• Valve operators that operate slowly safety-related system and performs the following safety-

• Limited reactor coolant system depressurization related function: using only the first stage ADS valves.

• Provide a controlled, sequenced depressurization of The automatic depressurization system valves the reactor coolant system. automatically actuate on receipt of a signal from the protection and safety monitoring system. This signal The automatic depressurization system, together with sequentially opens the ADS valves, with a time delay the passive core cooling system, provides core cooling between the actuation of each stage to prevent the during loss of coolant accidents, from the initial simultaneous opening of more than one stage at a time.

actuation of passive core cooling through the long-term The . ADS valves open when actuated and remain open cooling mode, for the duration of any event that requires ADS The automatic depressurization system consists of actuation.

four stages of depressurization valves. Each stage has The automatic depressurization system valves can two separate lines and each line has two valves in also be manually actuated in the main control room or h

V series. The four stages have a total of 16 valves. The at the remote shutdown workstation via the protection four stages open sequentially, with all four valves in and safety monitoring system or by dedicated manual cach stage opening simultaneously. The ADS also switches in the main control room via the diverse includes two spargers, located inside the in-containment actuation system. He protection and safety monitoring refueling water storage tank. system can be used to manually operate the ADS valves, The first, second, and third automatic either individually or via a system level actuation. The deptessurization system stages are arranged in two diverse actuation system provides the capability to parallel groups. Each group consists of three parallel manually actuate the valves in each ADS valve stage.

lines, one line for each first, second, and third ADS he valves that initiate automatic depressurization stage. Each group also includes a common inlet line system receive Class IE de power. The valves energize that connects to the top of the pressurizer and a common to accomplish their safety-related functions, discharge line that connects to a sparger. The automatic depressurization system There are two separate fourth stage automatic accommodates single failures of any individual ADS depressurization system lines, each connecting to one valve or any of the Class IE de power divisions.

reactor coolant system hot leg and venting to containment.

The automatic depressurization system provides a controlled reactor cooicnt system depressurization through the incorporatinn of the following features:

• Sequential opening of the valve stages

• A time delay between sequential stages

• Different valve sizes for the valve stages O 3.2.1 -1 V W Westinghouse

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Tier 1 Design Certification Material AUTOMATIC DEPRESSURIZATION SYSTEM W d Rwision: 1 Effective: 12/15/92 . ..

Table 3.2.1 Automatic Depressurization System inspections, Tests Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. The flow paths required to provide a inspections shall be conducted of the as-built a. The as-built configuration of the flow paths controlled, sequenced depressurization of automatic depressurization system. for the ADS is as follows:

the reactor coolant system are included in the as-built configuration for the automatic (1) Two inlet lines, each connecting a depressurization system. separate nozzle in the top of the pressurizer with the inlet to one of the two parallel first/second/ third stage valve groups.

(2) Tuo first/second/ third stage valve groups. each consisting of three parallel lines, one line for each of the associated ADS valve stages. The line for each stage contains two remotely-operated isolation valves in series.

(3) Two discharge lines, one from each first/second/ third stage valve group connecting to the associated sparger, located in the in-containment refueling

- water storage tank.

(4) Two fourth stage discharge lines, one from each reactor coolant system hot leg, discharging to containment. Each discharge line contains two remotely-operated isolation valves in series.

W - Westin=crhouse a

('~% p Tier 1 Design Certification Material AUTOMATIC DEPRESSURIZATION SYSTEM -

Flevision: 1 Effective: 12/15/92 _

i Table 3.2.1 Automatic Depressurization System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. (continued) b. De elevation of bottom of the in-containment refueling water storage tank.

overflow pipe is higher than the sparger, where the top of the arms connect to the central hub by:

s [10.5] feet

c. The elevation of each fourth stage discharge pipe outlet centerline is higher than the direct vessel injection nonle centerline by:

2 [12.0] feet 6 [14.0] feet

2. The automatic depressurization system a. Functional tests shall be conducted a. The following ADS valves open upon valves open tipon receipt of an actuation demonstrating valve operation upon receipt receipt of an actuation signal, with valve signal from the protection and safety of an actuation signal from the protection stroke times as specified:

monitoring system and the diverse actuation and safety monitoring system.

system. 2 seconds s seconds First Stage [20] [30]

Second Stage [40] [120]

Third Stage [90] [120]

Fourth Stage [10] [30]

W Westinghouse

v v Tier 1 Design Certification Material J

AUTOMATIC DEPRESSURIZATION SYSTEM -

Revision: 1 Effective: 12/15/92 Table 3.2.1 Automatic Depressurization System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections Tests, Analyses Acceptance Criteria

2. (continu d) b. Functional tests shall be conducted b. He following ADS valves open upon demonstrating valve operation upon receipt receipt of an actuation signrJ.

of an actuation signal from the diverse actuation system.

3. The disision assignment for valves and With only the assigned Class IE power and He ADS valves assigned to the energized controls is such that the loss of any single protection division energized, tests shall be division receive actuation signals and are Class 1E division will not prevent the conducted to confirm power and protection powered as follows:

automatic depressurization system from division assignments by operating the control performing its safety-related function of circuits. His test shall be performed for each Each of the two groups of Grst/second' third presiding a controlled, sequenced assigned power division. stage ADS valves are powered from two of the depressurization of the reactor coolant four Class IE de power divisions. Within each system. group, nrst and third stage are powered from the same division and second stage is powered from the other division.

He two fourth stage ADS lines are configured so that both valves in one line are powered from two Class 1E de power divisions, so that actuation of either division can open both valves. He other fourth stage ADS line is configured the same, using the other two divisions.

W

_ WestinEhouse

y ,

d d Tier 1 Design Certification Meteria!

b AUTOMATIC DEPRESSURIZATION SYSTEM 4 Revision: 1 Effective: 12115/92 ,

Table 3.2.1 Automatic Depressurization System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

4. The first, second, and third stage automatic A functional test shall be conducted to verify Each first, second, and third stage ADS valve depressurization system valves are capable that each Grst, second, and third stage ADS opens.

of opening with a differential pressure of valve can open with an initial differential

[2250] psid across the valve seat. pressure of at least [2250] psid across the valve seat.

5. The fourth stage automatic depressurization A functional test shall be conducted to verify Each fourth stage ADS valve opens.

system valves are capable of opening with a that each fourth stage ADS valve can open with differential pressure of [1000] psid across an initial differential pressure of at least the valve seat. [100f)] psid across the valve seat.

6. The automatic depressurization system a. Inspections shall be conducted to verify the a. The throat areas for the ADS valves are as valves provide the required vent How from throat areas of the ADS valves. follows:

the reactor coolant system.

Stage 2 square inches 5; square inches 1 [6] [8]

2&3 [28] [34]

4 [76] [86]

W-Westinehousea

P (l

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O\

Tier 1 Design Certification Material AUTOMATIC DEPRESSURIZATION SYSTEM >

Revision: 1 Effective: 12/15/92 _

Table 3.2.1 Automatic Depressurization System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

6. (continued) b. A low pressure vent flow test and associated b. He ADS piping flow resistance for the analysis shall be conducted to determine the specified ADS valve combinations is:

total piping flow resistance of each first/second!!hird stage ADS valve group Staces Or>en Total Resistance from the pressurizer to the sparger.

1&3 s [2.63E-6] ft/gpm 2 He reactor ecolant system shall be at cold conditions with the pressurizer full of water 2 s [2.81E-6] ft/gpm 2 and with the reactor coolant pumps stopped.

1&2&3 s [1.27E-7] ft/gpm 2 he normal residual heat removal pumps shall be used to provide injection flow into the reactor coolant system, discharging through the ADS valves.

c. Inspection and associated analysis of each c. (1) ne calculated flow resistance for the fourth stage ADS valve group shall be as-built fourth stage line routing is conducted to verify the line routing is consistent with the calculated design consistent with the line routing used for piping flow resistance:

design flow resistance calculations and to Staces Or.en Total Resistance verify that each line is free of internal obstructions. 4 s [9.74E-8] ft/gpm 2 (2) Each fourth stage line is free of obstructions.

W Westinghouse

Tier 1 Design Certification Material CONTAINMENT SYSTEM Revision: 1 9.*

Effective: 12/15/92 3.2.2 CONTAINMENT SYSTEM Design Description ne containment system (CNS) is a safety-related

• Isolation valves are designed to close with the system and is a barrier designed to prevent the release prevailing conditions that may exist during events of fission products to ti.e atmosphere. requiring containment isolation.

The containment system is formed by the steel containment shell, electrical and mechanical

• ne division assignment for valves and controls is penetrations, fuel transfer channel penetration, such that the loss of any single Class IE power equipment batches, personnel airlocks, steam generator division will not prevent containment isolation.

shells, and the steam, feedwater and blowdown lines within the containment. Although the containment

• No single failure prevents containment isolation.

isolation valves are not a part of the containment system, they perform an essential containment system Normally open mechanical penetrations are isolated function and are verified as part of the containment by the appropriate isolation signal. He actuation signal system. may be generated:

The containment system can withstand the highest internal pressure and temperatures resulting from a

• Automatically or manually within the protection and postulated loss-of-coolant accident (LOCA), main steam safety monitoring system at a system level line break or feedwater line break. The system also can withstand transient conditions from other postulated

• Automatically or manually within the dive se events including discharge of the passive containment actuation system at a system level cooling system (PCS) water or loss of ac power.

The containment system and the contMnment

• Manually at an individual valve level within the isolation valves provide the isolation function by protection and safety monitoring system.

establishing a barrier between the containment -

environment and the outside environment. Containment The manual protection and safety monitoring system isolation provisions isolate lluid lines which penetrate isolation signals can be initiated from either the main the primary containment boundary so as to mitigate the control room or the remote shutdown workstation.

release of radioactivity to the environment. He system With the passive containment cooling system, the incorporates the following features: containment system performs the function of heat remosal from the containment. Refer to the passive

• Automatic containment isolation valves are actuated containment cooling system Tier i Design Description by Class IE de power. in Section 3.2.3.

• Remotely operated non-motor-operated valves fail in the closed position upon loss of support system, such as instrument air or electric power.

• Normally closed manual containment isolation valves have provisions for locking the valve closed.

24 W Westinghouse

• ' - e .A Tier 1 Design Certification Material

=;

gg CONTAINMENT SYSTEM '"

Revision: 1 Effective: 12/15/92 Table 3.2.2 Containrt ent System inspections, Tests, Analyses and Acceptance Criteria inspections, Tests. Analyses Acceptance Criteria Certified Design Commitment

?_ containment pressurization test shmil be The centainment vesel has an N

i. The containment system is constructed with conducted. Symbol Stamp a design pressure of 45 psig.

Valve functional tests shall be conducted a. Valves close consistent with the

2. The remcc operated containment demonstrating proper valve operation upcn isolation criteria of the containment isolation valves close upon receipt of receipt of each appropiate actuation signal. system Itincipal Co.stainment automatic or manual closure signals from Penetrations Table 3.2.2-2 upon the protection and safety mocitoring system receipt of a protection and safety or diverse actuation system.

monitoring sys:em signal.

b. Va!ves close upon receipt of a diverse actuation signal.

With only the assigned Class iE power and No penetration depends on a single

3. The division assignment for valves and protection division energized. a test shall be electrical power division for the controls is such that the loss of any single conducted + > confirm power and prctection successful isolation.

Class iE division will not prevent division assignm.nts by operating the control containment isolation.

circuit. This test shall be performed for each assigned power division.

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.2.2 2 W_

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Tier 1 Design Certificat;on Material CONTA!NMENT SYSTEM l Revision: 1

- Effective: 12/15/92 Table 3.2.2 Containment System

' Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment Inspections. Tests, Analyses Acceptance Criteria

4. The containment system and the isolation - A test program shall be conducted to determine "Ihe overn!! integrated leakage rate as' f valves are an effective leak-tight barrier the containment system overall integrated determiw. d and confirmed is less than against an uncontrolled release of leakage rate at a containment pressure of design leakage rate of [.12] weight l

radioacdvity to th: environment. 45 psig. percent of contained air mass at test

pressure per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.  ;

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Tier 1 Design Certification Material CONTAINMENT SYSTEM .))

Revision: 1 51 Effective: 12/15/92

. -s l Table 3.2.2 PRINCIPAL CONTAINMENT PEr4ETRAT10NS Device ladon Maximum Device cadon Nnedon .9" System Line Description (3) e CAS Breathing Air In IRC Containment isolation None N/A ORC Containment imiation None N/A Service Air in IRC Containment Imlation None N/A ORC Containmmt Isolation PS15 [601 mconds g CCS CCW IRC *.eads in IRC Containment Isolation PS15 160] mconds ORC Containment isolation PN1S [60} econds CCW IRC Loads Out IRC Thermal Relief None N/A IRC Containment isolation PN15 [60] secends ORC Containment Isolatimi PNis [60] seconds CVS Spent Resin Flush Out IRC Dermal Relief Nene N'A IRC Centainment Imlation None N/A ORC Centainment imlation None N!A Letdow n IRC Thermal Relief Nene N/A IRC Containment isolation PNIS (60] seconds ORC Containment isolation PSIS [60] weonds W Westinghouse

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Tier 1 Design Certification Material CONTAINMENT SYSTEM Revision: 1 Effective: 12/15/92 Table 3.2.2 PRINCIPAL CONTAINMENT PENETRATIONS Isolation Maximum Device Device , al ute Location Function System Line Description IRC Containment Isolation Pats {t0] seconds Charging ORC Containment Isolation Pats [60] seconds IRC Containment Isolation None N/A H Injection to RCS ORC Containment Isolation Pats [60} seconds IRC Containment Isolation Nene N/A Water to CMT & Accumulators ORC Containment Isolation P%fS [60J seconds IRC Containment Isolatien None N/A DWS Demineralized Water System ORC Containment Isolation PNTS [60] seconds IRC Containment Isolation None N/A Fl{S Fuel Transfer IRC Containment Isolation None N/A FPS- Fire Protection Standpipe ORC Containment Isolation None N/A IRC Containment Isolation None N/A PCS Containment Pressure ORC Containment Isolation None N/A IRC Containment Isolation None N/A Containment Pressure ORC Containment isolation None N/A

.2.2.s w westinghouse

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O O Tier 1 Design Certifiestion Material O

CONTAINMENT SYSTEM y Revision: 1 Effective: 12/15/92 Table 3.2.2 PRINCIPAL CONTAINMENT PENETRATIONS

'* ladon Madmum Device Device .

C* " "" "

System Line Description (1)

Containment Pressure IRC Containment Isolation None N/A ORC Containment Isolation None NfA Containment Pressure IRC Containment Isolation None N 'A ORC Containment Isolation None N/A PSS RCS/PXS'CVS Samples Out IRC Containment Isolation PNiS [60] seconds IRC Containment Isolation P5tS [60] seconds ORC Containment Isclation PSIS [60] seconds Containment Air Samples Out IRC Containment Iselation PhtS [60] seconds IRC Containment Isolation P5tS [60} seconds ORC Containment Iselation PSIS [60] seconds RCS/ Containment Air Sample Return IRC Containment Isolation None N/A ORC Containment isolation PhtS [60} seconds Spare IRC Containment Isolation None N/A ORC Containment Isolation g None N/A W Westinghouse

i i

) a J a Tier 1 Design Certification Material y

CONTAINMENT SYSTEM Revision: 1 Effective: 12/15/92 Table 3.2.2 PRINCIPAL CONTAINMENT PENETRATIONS _

Isoladon Madmum Device Device ignal 1 sure Location Function #1) Time Line Description System _ _ . ~e IRC Le N/A PXS N: to Accumulators Containment Imlatier [I ORC Con".ainment Isciatb, ! cfS [60] seconds IRC RCS Hot leg i<4 t;. '

.ne N/A RNS RCSflRWST to RHR Pump IRC N'A 1 RCS Hot tzg lu satinn ,o _ " .one IRC IRWST Une Isolation g PSIS [60! seconds l

IRC PXS Line Isolation None N/A .

IRC CVS Line isola: ion None N'A IRC Relief to IRWST None N/A ORC Containment Isolation None N/A IRC Containment Isolation None N/A RHR Pump to RCS ORC Containment Isolation None N/A IRC Containment isolstion None N/A SFS SF Pump to IRWST/Ref. Cavity ORC Centainment Isolation PS!S [60j seconds 3.2.2-7 V_v, Yiestinghouse f

p /'h Tier 1 Design Certification Material CONTAINMENT SYSTEM jg gj Revision: 1 "i Effective: 12/15/92 Table 3.2.2 PRINCIPAL CONTAINMENT PENETRATIONS is I tion Maximum Deh Dde System Line Description (1) e IRWST/Ref. Cavity Purif. Out IRC Thermal Relief None N/A IRC Containment Isolation PNIS [60] reconds ORC Containment Isolation PMS [60] second, SGS Main Steam line #1 ORC Main Steam Isclation PMS {5] seconds ORC Main Steam Bypass PNIS {l0} seconds Iwlatien ORC SG PORV Imlatien PMS [10] m nds ORC Condennte Drain Isol. PMS [IOl seconds h ORC MS Safety Valve None N 'A ORC MS Safety Valve None N/A ORC MS Safety Valve None N/A Main Steam line #2 ORC Main Steam Isolation PMS [5] seconds ORC Main Steam Bypass PMS [loj secends twlation ORC SG PORV twlation PMS {l0} wonds

.2.2-8 W Westinghouse

r O Tier 1 Design Certification Material O

CONTAINMENT SYSTEM y E]

Revision: 1 "'

El Effective: 12/1'i/92 Table 3.2.2 PRIT4CIPAL CONTAINMENT PEN.ETRATIONS Isoladon Madmum Device Device Location Function

,egnal Cosure System L.me Desen.ption (1) Time ORC Condensate Drain Isol PMS [10] seconds ORC MS Safety Valve None N/A ORC MS Safety Valve None NTA ORC MS Safety Valve None N/A Main and Gtartup Fecowater #1 ORC Main Feed Isolation PMS [5] seconds ORC Startup Feed Isolation PMS [10] seconds Main and Startup Feedwater #2 ORC Ma;n Feed Isolation PMS {5] seconds ORC Stntup Feed Isolation PMS [10] seconds SG Blowdown #1 ORC Centainment isolation PMS [10} secends SG Blowdown ,",2 ORC Containment Isolation PMS [10] seconds SG Blowdown Recirculation ORC Containment Isolation None N /A VFS Containment Air Filter Supply A IRC Containment Isolation PMS.DAS [5] seconds ORC Containmere. Isolation PMS.DAS [5] seconds Containment Air Filter Supply B IRC Containment Isolatien PMS.DAS [5] w onds ORC Containment Isolation PMS.DAS [5] seconds (vV Westinghouse

O O Tier 1 Design Certification Material CONTAINMENT SYSTEM jg Revision: 1 "

Effective: 12/15/92 Table 3.2.2 PRINCIPAL CONTAINMENT PENETRATIONS Isoladon Madmum Device Device ,

System Line Descript on i (1 e Containment Air Filter Exhaust A IRC Containment Isolation P51S.DAS [5] seconds ORC Containment Isolation P515.DAS [5] seconds Containment Air Filter Exhaust B IRC Containment Isolation P5tS.DAS [5] seconds ORC Containment isolation P515.DAS [5] seemds VWS Chilled Water from Fan Coolers IRC Containment isolation P5tS [60] seconds ORC Containment isolation P51S [60] seconds Chilled Water to Fan Coolers IRC Containment Isolation PAIS [60] seconds ORC Containment Isc,lation P515 [60] seconds WLS Reactor Coolant Drain Tank Out IRC Containment isolation P51S.DAS [60] seconds ORC Containment isolation P\lS.DAS [60] seconds Reactor Coolant Drain Tank Gas IRC Containment Isolation Pats [60] seconds ORC Containment Isolation PSIS [60] seconds Normti Containment Sump IRC Containment Isolation PAIS.DAS [60] seconds ORC Containment isolation Pats.DAS [60] seconds

.2.2 20 w wainocuse

'N /'N Tier 1 Design Certification Material CONTAINMENT SYSTEM gm Revision:.1 Effective: 12/15/92 _

Table 3.2.2 PRINCIPAL CONTAINMENT PENETRATIONS is la n admum Device Device gnal C osure

. Location Function 3 System L.me Description (1) Time SPARE Spare IRC Containment Isolation Nor,e N'A ORC Containment Isolation None N/A CNS Main Equipment Hatch IRC Containment Isolation None N/A

, Maintenance Hatch IRC Containment Isolation None N/A Personnel Hatch IRC Containment Isolation None N/A .

ORC Containment Isolation None N/A Personnel Hatch IRC Containment Isolation None N/A ORC Containment Isolation None N/A f Electrical Penetrations Containment isolation None N/A i

NOTES:

1) Penetratien signal applies to automatic remote isolation valves for the penetra: ion. r 1

PMS: Prttection and Safety Monitoring System Signal DAS: Diverse Actuation System Signal l'

r 3.2.2-11 3 Westinghouse 1

TI:r 1 D: sign Certific: tion Mit:rlat m

I T PASSIVE CONTAINMENT COOLING SYSTEM V Revision: ? , NI Effective: 12/15/92 3.2.3 PASSIVE CONTAINMENT COOLING SYSTEM Design Description 1

1he passive containment cooling system (PCS) is a 1he passive containment cooling system performs  ;

safety-related system w hich reduces the containment the following safety-related functions. j temperature and pressure following a loss-of-coolant I accident or a main steam line break inside containment.

• The PCS provides containment heat removal l 1he passive containment cooling system serves as the capability following a postulated design basis means of transferring heat to the atmosphere for design accident such as limiting loss-of-coolant accidents basis events which result in a significant increase in and secondary side accidents (main steam line break containment pressure and temperature. 1he passive or feedwater line break). 1he PCS removes containment cooling system has the capability to remove containment heat to limit containment peak pressure energy from the reactor containment structure to prevent to less than the design pressure.

the containment from exceeding its design pressure and to reduce containment pressure following design basis

• 1he PCS reduces containment pressure to less than es ents. The safety-related portions of the passive one half of containment design pressure within containment cooling system include: 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a postulated loss-of coolant accident.

7

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• Passive containment cooling water storage tank, b delivery lines and associated valves up to the * 'lhe PCS provides containment heat removal for a containment ve.sel water distribution bucket period of at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with no operator actions or outside assistance such that the containment

• Passive containment cooling water storage tank vent design pressure will not be exceeded.

lines

• 'the PCS provides continued containment cooling

• Containment vessel water distribution system water flow via a temporary safety-related cooling water connection following water storage tank

• Shield building annulus drains depletion after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

• PCS cooling air flow path formed within the shield 1he passive containment cooling system performs building comprised of the air inlets, outer annulus, the following defense-in-depth functions:

inner annulus, diffuser area and air / vapor exhaust structure

• The passive containment cooling water storage tank provides a water source for the fire protection

• Passive containment cooling water storage tank system. Refer to the fire protection system Tier 1 flanged makeup connection from the flange to the Design Description in Section 3.3.4.

storage tank.

The passive containment cooling system transmits j heat directly from the reactor containment structure to

the environment such that the containment design pressure (and temperature) are not exceeded following

/

3.2.3 1 (s W-Westinghouse L

Ti:r 1 D: sign Certific: tion M:t: rial O PASSIVE CONTAINMENT COOLING SYSTFM Revision: 1 Effective: 12/15/92

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'!he manual protection and safety monitoring system signals can be initiated frorn either the main control room of the remote f.hutdown workstation.

O d

W Westinghouse

m ~. A Tier 1 Design Certification Material PASSIVE CONTAINMENT COOLING SYSTEM y Revision: 1 Effective: 12115/92 l-Table 3.2.3 Passive Containment Cooling System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspection, Test, Analyses Acceptance Critaria

1. The PCS provides water delivery to the a. Tests of the PCS water storag- tank delivery a. With the PCS sterage tank at the specified containment vessel for cooling following rates shall be conducted to verify system u.ater level and with one pair of isolation design basis accidents. flow performance from each potential flow v:Ives open. the water flow delivery for path delivering separately. each flow path is 2:

• [220] gpm at tive minimum operating PCS storage tank les el i [2] inches

• [o5] gpm at [TBD] abose the intermediate standpipe [2] inches

= [50] gym at [12] inches i [2] inches above tank bottom outlet nozzle.

b. A test of the PCS water sierage tank b. With the PCS storage tank at the specified delivery rates shall be conducted to verify water level and with all isolation s alves system flow performance with both flow open, the water flow delivery for each flow paths deliv. ring simultaneously. path is s:

• [2N] gpm at the minimum operating PCS storage tank level i [2] inches

+ [105] gym at [TBD] abose the intermediate standpipe [2] inches

= [60] gpm at [12) inches [2] inches above tank exhaustion W Westinghouse

O O Tier 1 Design Certification Material PASSIVE CONTAINMENT COOLING SYSTEM _- =i; Revision: 1 Effective: 12115/92 _

E Table 3.2.31 - Passive Containment Cooling System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspection, Test, Analyses Acceptance Criteria

2. He PCS discharge isolation and block a Functional tests shall be conducted a. Each PCS isolation valve and the block

' valves open upon receipt of the specified demonstrating proper valve operation of valve (when closed) opens.

, actuation signal from the protection and each isolation and bkxk valve upon receipt safety monitoring system or diverse of an actuatien signal from the protection actuation system. and safety monitoring system.

b. Functional tests shall be conducted b. Each isolation valve opens.

demonstrating proper valve operation of each isolation valve upon receipt of an actuation signal from the diverse actuation system.

3. He division assignment for isolation vahes With the assigned Class 1E power and %e PCS valves assigned to the energized and controls is such that the loss of any protection division energized, tests shall be division receive an actuation signal and have single Class 1E division does not prevent conducted to confirm power and protection power available.

the PCS from accomplishing its safety- division assignments by operating the control related function of proCding containment circuits. His test shall be performed for each heat removal & reducing containment assigned power division.

pressure.

4. A makeup flow path to the PCS storage A test of the capability to provide cooling uater Makeup flow can be delivered to the PCS water  !

tank provides for continued PCS cooling flow directly to the PCS sterage tank shall be storage tank is 2 [55] gpm.

flow. conductcd from the PCS makeup flanged connection. i i

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Tier 1 Design Certification Material t

{

PASSIVE CONTAINMENT COOLING SYSTEM 1 Revision: 1 Effective: 12/15/92 s

Table 3.2.3 Passive Containment Cooling System Inspections, Tests, Analyses and Acceptance Criteria ,

Certified Design Commitment Inspection, Test, Analyses Acceptance Criteria

5. He water volume in the tank provides PCS Tests or as-built dimensions and volumetric Water volumes for the following tank levels cooling water flow for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and fire analysis will be performed to terify that the are:

protection system supply requirements. required water volumes are prmided.

• 2 [400.000) at the fire protection system / recirculation entrance nozzle .

elevation

• 2 [TBD] gallens between the high and the intermediate PCS delivery standpipes

• 2 [TBD] gallons below the intermediate PCS delivery standpipe

• 2 [6000] gallons between the tank overflow and fire protection system recirculation entrance noz2le f

elevation

6. The annulus drains provide for drainage of A test shall be conducted to confirm the The annulus drainage rate is 2 [230] gpm with excess containment cooling water that has capability of the annulus drains. an annulus water level of s [12] inches.

not evaporated. ,

7. The PCS provides for effective containment A test shall be conducted to demonstrate he containment vessel circumference is 2 shell wetting. effective vessel wetting at the minimum cooling [35]% wetted as measured at the upper spring  !

flow rate. line with a minimum water Mel in the PCS  ;

storage tank. of [12] inches i [2] inches above j the tank outlet nozzle with a single flow path open. 1 W Westinghouse '

Ti:r 1 D: sign Ccrtification M:t: rial (nv) PASSIVE CORE COOLING SYSTEM Revision: 1 h Effective: 12/15/92 3.2.4 PASSIVE CORE COOLING SYSTEM Design Description The passive core cooling system (PXS) performs the The subsystems that provide the passive core cooling following safety related functions: functions are:

• Emergency Core Decay lleat Removal

• Passive residual heat removal heat exchangers

• Core makeup tanks Provide core decay heat removal whenever the

• Accumulators normal heat removal paths are disabled.

• In-containment refueling water storage tank and containment recirculation

• Emergency Reactor Coolant System Makeup and a pH adjustment tank.

lloration The passive core cooling system design description Provide reactor coolant system makeup and boration is divided into sections, based on the individual when the normal reactor coolant system makeup subsystems that comprise the passive core cooling supply from the chemical and volume control system system. Each subsystem has a separate design is unavailable or is insufficient. description.

• Safety injection 3.2.4.1 Passive Residual Heat Removal Heat Provide safety injection to the reactor coolant system Exchangers to provide core cooling for the complete range of loss-of-coolant accidents, up to and including the lhe passive residual heat removal heat exchangers double-ended rupture of the largest reactor coolant provide safety-related core decay heat removal whenever system piping. normal reactor coolant system hat remosal paths are not available.

• Containment pH Control The passive residual heat removal heat exchanger subsystem includes two passive residual heat removal Provid- for chemical addition to the containment heat exchangers, located inside the in-containment during post-accident conditions to establish floodup refueling water storage tank, and the associated valves chemistry conditions that support radionuclide and piping that connect the heat exchangers to the retention in the event of high radioactivity in reactor coolant system, containment. The passive residual heat removal heat exchanger subsystem has a common inlet line from one of the The passive ccee cooling system depends upon reactor coolant system hot legs with remotely-operated passive components and processes such as gravity isolation capability. The common inlet line is connected injection and expansion of compressed gases. 1he to two parallel heat exchangers. A common outlet line passive core cooling system requires a one-time connects to the steam generator cold leg channel head alignment of valves upon actuation of the specific on the same reactor coolant system loop as the heat colnponents.

4

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Ti:r 1 D: sign Certific tion M:ttrizi PASSIVE CORE COOLING SYSTEM N Revision: 1 Ef fective: 12/1!i/92 exchanger inlet piping. 1he outlet line contains two connect the core makeup tanks to the teattor coolant parallel isolation valves. system.

1he passisc residual heat removal heat exchangers liach core makeup tank has an inlet pressure balance actuate on receipt of a signal from the protection and line from the pressuriter with a remotely-operated t.afety momtoring splem or from the diverse actuation isolation valve and two series check valves, and a inlet system. I;ither automatic actuation rignal opens the prenure balance line from a reactor coolant system cold outlet isolation valves for the passive residual heat leg. I!ach cold leg prenure balance line has two removal heat exchanrers. The protection and safety parallel, remotely-operated isolation valves.11ach CMT monitoring system also provides an open signal to the has an outlet discharge line that connects to a direct common inlet line isols. tion valve. venel injection line, w hich injects into the reactor veuel lhe passive residual heat removal heat exchangers downeomer liach core makeup tank discharge line can also be manually actua:ed in the main control room contains two parallel remotely-operated isolation valves, or at the :cmote shutdown workstation via the proicction two series check valves, and a flow tuning orifice, and safety monitoring system or by dedicated switches lhe core makeup tanks actuate on receipt of a signal in the nusin centrol room via the diverse actuation from the protection and safety monitoring system or sptem. The protection and safety monitoring system from the diverse actuation system. liither automatic can be used to manually operate the inlet and outlet actuation signal opens the inlet and outlet isolation isolation valves, either individually or via a sptem level valves for each core makeup tank. The protection and actuation. 'ihe diverse actuation sptem prmides a safety monitoring system also provides an open signal to system level actuation ,gnrJ for the outlet isolation the isolation valve in each preuuriter pressure balance valves. line.

The valses that initiate panive residual heat removal The core makeup t:nks can alm be manually heat exchanger injection receive Class III de power. actuated in the main control room or at the remote

'the outlet valves de-energite to actuate their safety- shutdown workstation via the protection and safety related functions.1he common inlet isolation valve is monitoring sptem or by dedicated switches in the main normally open, and it receises an actuation signal to control room via the diverse actuatbn sptem. The open from the protection and safety monitoring sptem. protection and safety mon..oring system can be used to manually operate the cold leg inlet line, the pressuriter 3.2.4.2 Core Makeup Tanks inlet line, and the outlet line isolation valves, either individually or via a system level actuation. The diverse The core nutkeup tanks provide safety related actuation system provides a sptem level actuation signal emergency makeup to the reactor coolant system w hen for the cold leg inlet line and the outlet line isolation the nonnal makeup sptems are not available.1he core valves.

makeup tsnLs are also one of the sources of passive The valves that initiate core makeup tank injection safety injection available during loss of coolant receive Cla*s 111 de power. The inlet and outlet valves accidents, with the other sources being the de-energire to actuate their safety-related functions.

accumulators, the in-containment refueling water storage When the core makeup tanks actuate, the inlet line tant, and the containment recirculation , from the pressurizer supplies steam to allow core lhe core makeup tank subsystem includes two core makeup tank injection to mitigate non-LOCA events, makeup tanks and the associated valves and piping that 1he large inlet line from the cold leg is sited for loss of coolant accidents, where higher core makeup . tank injection flows are required, 3.2.42 W- Westinnhouse a

Ti:r 1 D: sign C:rtific: tion M:t: rid

[

PASSIVE CORE COOLING SYSTEM Revision: 1 g Effective: 12/15/92

. e 3.2.4.3 Aci itors ne in-containment refueling water storage tank and containment recirculation subsystem includes the in-He acc umt . ors are one of the safety-related containment refueling water storage tank, two sources of passive safety injection available during lon containment recirculation screens, and the associated of coolant accidents. vahes and piping that connect the tank and the screens

%e accumulator subsystem includes two to the reactor coolant system.

accumulators and the associated salves and piping that ne IRWST has two injection lines, each of which connect the accumulators to the reactor coolant system. unnects to a direct vessel injection line, which inject lxh accumulator has a discharge line that connects into the reactor vessel downcomer. Each IRWST to a direct venel injection line, which injects into the injection line contains an isolation valve. a now tuning reactor vessel downeomer. Each accumulator discharge orifice, and four check valves (two parallel paths of two line contains a discharge isolation valve, two series series check valves).

che<.k valves, and a How tuning orifice. Each containment recirculation screen has two lines The accumulators automatically provide injection to which join together before they connect to an IRWST the reactor coolant system when system pressure falls injection line, upstream of its isolation valves. One of below the static pressure in the accumulator and the the two parallel recirculation lines from each screen discharge check valves open. He remotely-operated contains two series check valves and the other line discharge isolation valve for each accumulator is contains two series isolatiot' valves.

normally open, but it receives an actuation signal trom %e in-containment refueling water storage tank the protection and safety monitoring system, automatically provides How through each injection line ne accumulator discharge isolation valves can also after the reactor coolant system has been depressurized be manually actuated in the main control room or at the sufficiently, so that the head of water in the tank can remote shutdown workstation, either individually or via open the IRWST injection line check alves. Each a system level actuation by the protection and safety IRWST injection line has a normally open, remotely-monitoring system. operated isolation valve which receives an open The accumulator discharge valves receive Class IE actuation signal from the protection and safety de power. Dey are normally open and are not required monitoring system and the diveae actuation system.

to reposition to initiate accumulator injection flow. Containment recirculation initiates automatically through either the check valve or the remotely operated valve pwn. He check valve paths open when the 3.2.4.4 In-Containment Refueling Water containment 0oodup elevation exceeds the lRWST ies el, Storage Tank and Containment The remotely operated valve recirculation now path Recirculation pens automatically on a signal from the protection and safety monitoring system.

He in-containment refueling water storage tank The injection line and the recirculation line remote (IRWST) and containment recirculation are two of the imlation valves can be manually actuated in the main safety-related sources of panive safety injection e ntr i r m or at the remote shutdown workstation via available during loss of coolant accidents. Dese sources the protection and safety monitoring system. either preside injection and recirculation following reactor individually or via system level actuations.

coolant system depressurization during loss of coolant The valves that remotely actuate to initiate injection accidents. After containment Hoodup, long term core and recirculation How receive Class IE de power.

cooling is provided by containment recircalation.

O 3.2.4-3 V W Westinghouse

Ti:r 1 D: sign Certific: tion M:t: rial PASSIVE CORE COOLING SYSTEM 1" lip Revision: 1 Effective: 12/15/92 e

3.2.4.5 The pH Acfjustment Tank The pl{ adjustment tank provides the safety-related addition of a pil control agent to the containment in certain accident floodup conditions where core damage has occurred and core radioactivity has been released from the reactor coolant system into containment.

The pit adjustment tank subsystem includes the pil adjustment tank and the associated valves and piping that allow the pil adjustment tank to drain to the containment.

The pit adjustment tank inlet line contains twa parallel vacuum breakers. 'the pil adjustment tant discharge now path contains two parallel isolation valves. The discharge Dow path is cross connected downstream of the isolation valves. The path then divides into two lines providing flow to both containment screen areas.

The pit adjustment tank actuates on receipt of a signal from the mctection and safety monitoring system.

1he actuation signal automatically opens the discharge isolation valves. When the pit adjustment tank actuates, the vacuum breakers on the tank inlet line provide air flow into the tank allowing it to drain to the containment.

The pil adjustment tank can also be manually actuated in the main control room or at the remote shutdown workstation via the protection and safety monitoring system by actuating the discharge isolation valves, either individually or via a system level actuation.

The valves that initiate pH adjustment tank flow receive Class IE de power.

3.2.44

W Westinghouse i

i

Tier 1 Design Certification Material 4

PASSIVE CORE COOLING SYSTEM Revision: 1 Effective: 12/15/92 1

4

! Table 3.2.4-1.1 - Passive Core Cooling System (PRHR Heat Exchangers) inspections, Tests Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria 1

l. He flow paths required to provide inptions shall be conducted of the as-built a. De as-built configuration of the flow paths j emergency core decay heat removal are passive residual heat removal heat exchanger for the passive residual heat removal heat included in the as-built configuration for the subsystem. exchanger subsystem is as fellows:

i passive residual heat removal heat exchanger subsystem. (1) A common inlet line frem one of the reactor coolant system hot legs with I remotely-operated isolation valve.

(2) Two parallel heat exchangers.

(3) A common outlet line that connects to the steam generator cold leg channel head of the same reacter coolant system loop as the heat exchanger inlet piping. He outlet line contains two pam!!el iwlation valves.

b. he elevation of the passive residual h-at removal heat exchanger upper channel head centerline is higher than the hot leg piping top (outside s*2rface) by:

2 [26.3] feet l

+

R

?[} Westinghouse i

~ s Tier 1 Design Certification Material PASSIVE CORE COOLING SYSTEM i=

(3 .E"1 l

• Revision: 1 Effective: 12/15/92 i

Table 3.2.4-1.1 - Passive Core Cooling System (PRHR Heat Exchangers)

Inspections. Tests. Analyses and Acceptance Criteria '

Certified Design Commitment inspections. Tests, Analyses Acceptance Criteria j

1. (continued) c. He elevation of the passive residual heat removal heat exchanger upper channel head j t centerline is higher than the lower channel l

, head by:

2 [17.5] feet

2. He passive residual heat removal heat a. Functional tests shall be conducted a_ De heat exchang-r inlet iselatien salve and exchanger inlet and outlet isolation valves demonstrating proper valve operation upon outlet isciation valves open upen receipt of ,

open upon receipt of an actuation signal from receipt of an actuation signal frora the an actuation signal.

the protection and safety monitoring system pretection and safety monitoring system.

and the diverse actuation system.

b. Functional tests shall be conducted b. He heat etchanger outlet isolation vahes demonstrating proper valve eperation upon open upon receipt of an actuation sicnal.

receipt of an actuation signal from the diverse ,

actuation system.

, 3. He division assignment for valves and With only the assigned Class !E power and The heat exchanger outlet isolation valve controls is such that the loss of any single protection division energized. tests shall be assigned to the energized division receive an Class IE division will not prevent the passive conducted to confirm power and protection actuation signal and have power available.

residual heat removal subsystem from division assignments by operating the control performing its safety-related function of circuits. His test shall be performed for each t emergency core decay heat removal. assigned power dhision.

I i

3.2.4-6 W-Westinghouse  :

i l

t

Tier 1 Design Certification Material PASSIVE CORE COOLING SYSTEM Revision: 1 ,

Effective: 12/15/92 Table 3.2.4-1.1 - Passive Core Cooling System (PRHR Heat Exchangers)

Inspections, Tests, Analyses and Acceptance Criteria inspections, Tests, Analyses Acceptance Criteria Certified Design Commitment A high pressure heat removal performance test De total heat transfer rate frem each heat

4. Each passive residual heat removal heat for each passive residual heat removal heat exchanger operating individually is:

exchanger provides the required reactor coolant system heat removat exchanger shall be conducted to determine the heat transfer rate from the heat exchangers. 2 [1.3 E 8] BTU /hr s [I.8 E 8] HT11%r De reactor coolant system sha11 be at hot j standby conditions with the het leg temperature between [550]' and [560]'F with the reactor coolant pumps stopped. De in-containment refueling water storage tank (IRWST) water level shall be at least [2SJ feet above tie tank bottom and the water temperature between

[ 30]'F and [70]'F.

I W Westinghouse

. _ _ _ _ _ - E

Tier 1 Design Certification Material

!a PASSIVE CORE COOLING SYSTEM M Revision: 1 Effective: 12/15/92 Table 3.2.4-1.2 - Passive Core Cooling System (Core Makeup Tanks)

Inspections, Tests Analyses and Acceptance Criteria inspections, Tests, Analyses Acceptance Criteria Certified Design Commitment

a. The as-built configuration of the flow paths I. The flow paths required to provide safety inspections shall be conducted of the as-built for each core makeup tank is as follows:

injection and emergency reactor coolant core makeup tank subsystem.

system makeup and boration are inchided in An intet pressure balance line from (1) the as-built configuration for the core the pressurizer with a renntely-makeup tank subsystem. operated isolation valve and two wries check valves.

C) An intet pressure balance line from one of the reactor coolant system ecid legs with two parallel, remotely-operated inlet isolation valves.

O) An outlet discharge line that connects to one of the direct vessel injection lines, which injects into the reactor vessel downcomer. Each core r:mkeup tank discharge line contains two paraIIel remotely-opersted iwlation valves, two series check valves, and a flow tuning orifice.

i 3.2.4-8 l l

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, ...._._m _ . _ _ . .

O '

Tier 1 Design Certification Material O

PASSIVE CORE COOLING SYSTEM Revision: 1 Effective: 12/15/92 Table 3.2.4-1.2 - Passive Core Cooling System (Core Maket1p Tanks)

Inspections, Tests, Analyses and Acceptance Criteria  ;

I inspections. Tests, Analyses Acceptance Criteria Certified Design Commitment  !

l. (contimmed) b. De elevation of the core makeup tank top  ;

(outside surface) is higher than the dirret vessel injection nozzle centerime by:

a [28.5] feet

2. He core makeup tank cold leg pressure a. Functional tests shall be conducted a. He pressurizer pressure balance line balance line isolation valves. the outlet demonstrating proper valve operation upon isolation valve the cold leg pressure balance discharge isolation valves, and the receipt cf an actuation signal from the line isolatien salves. and the discharge pressurizer pressure balance line isolation prctection and safety monitoring system. isolation valves open upon eceipt of an valve open upon receipt of an actuation actuation signal.

signal frem the protection and safety monitoring system and the diverse actuation b. Functional tests shall be conducted b. De cold leg pre =sure balance line isolation system. demonstrating proper valve operation upon valves and the discharge isolation valves receipt of an actuation signal from the diverse open upon receipt of an actuation sigral.

actuation system.

J

3. De division assignment for vrJves and With only the assigned Class IE power and He cold leg pressure balance line and controls is such that the loss of any single protection division energized. tests shall be discharge line isolation valves assigned to the Class 1E division will not prevent the core conducted to confirm power and protection energized division receise an actuation signal makeup tank subsystem from performing its division assignments by operating the centrol and have pcwer available.

safety function of providing safety injection circuits. This test shall be performed for each and emergency reactor coolant system assigned power division.

makeup and boration.

W Westinghouse i

i

O 3 Tier 1 Design Certification Material O

PASSIVE CORE COOLING SYSTEM ai .

~

Revision: 1 Effective: 12/15/92 ..

Table 3.2.4-1.2 - Pasdve Core Cooling System (Core Makeup Tanks)

Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

4. Each core makeup tank provides the required a. A low pressure injection test for each core a. The water injection flow rate from each core injection flow to the reactor coolant system. makeup tank shall be conducted to determine makeup tank, with the tank water level the CMT injection flow capability with air- between [22.90) and [23.10] feet above the compensated injection. direct vessel injection nozzle centerline, is:

The reactor vessel head and intemals shall be 2 [740}gpm removed, the reactor coolant system shall be s [835} gym empty, and the pressurizer and cold leg pressure balance line isolation valves shall be open.

Each core makeup tank shall be filled with water and the tank level change shall be used to determine the injection flow rate.

W Westinghouse

O O Tier 1 Design Certification Material O

PASSIVE CORE COOLING SYSTEM Revision: 1

{

~

Effective: 12/15/92 Table 3.2.4-1.2 - Passive Core Cooling System (Core Makeup Tanks) j inspections. Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

4. (continued) b. A low presmre test shall be conducted for b. De piping flow resistance of the core each CMT to determine the pressurizer makeup tank subsystem presmrizer pressure pressure balance line piping flow resistance. balance line piping for each core makeup tank is:

De reactor coolant system shall be at cold conditions with pressurizer full of water and 215.8 E -4] ft!gpm2 with the reactor coolant pumps stepped s [1.2 E -31 ft!gpm 2 De CMT discharge line iselation valves and the cold leg pressure balance line isolation valves shall be closed. De pressurizer pressure balance line isolation velve shall be open. Each CMT shall be initially drained and maintained vented to the containment.

The normal residual heat removal pumps sha!! be used to provide makeup flow from the IRWST into the reactor coolant system, discharging to the CMT.

W Westinghouse

O O Tier 1 Design Certification Material O

PASSIVE CORE COOLING SYSTEM in Revision: 1 E 4 Effective: 12/15/92 Table 3.2.4-1.2 - Passive Core Cooling System (Core Makeup Tanks)

Inspections. Tests. Analyses and Acceptance Criteria Certified Design Commitment inspections Tests. Analyses Acceptance Criteria

4. (continued) c. A low pressure test shall be conducted for c. He piping flow resistance of the core each CMT to determine the cold leg pr-ssure makeup tank subsystem cold leg pressure balance line piping flow resistance. balance line piping for each core makeup tank is:

De reactor coolant system shall be at cold conditions with pressurizer full of water and /

2 [6.9 E -6] ft gpm with the reactor coolant pumps stopped. f s; [9.3 E -6] ft gpm De CMT discharge line isclation valves and the pressurizer pressure balance line isolation valve shall be closed. De cold leg pressure balance line isolation valves shall be open.

Each CMT shall be initially drained and maintained vented to the containment.

The normal residual heat removal pomps shall be used to provide makeup flow from the IRWST into reactor coolant system, discharging to the CMT.

5. The core makeup tank volume shall be Tests or as-built dimensions and volumetrie ne solume of each core makeup tank is sufficient to provide the required injection. analyses shall be performed to verify the tank 2 [2000] cubic feet.

volume.

hV Westinghouse

O O Tier 1 Design Certification Material O

PASSIVE CORE COOLING SYSTEM Revision: 1 Effective: 12/15/92 .

Table 3.2.4-1.3 - Passive Core Cooling System (Accumulators)

Inspections, Tests Analyses and Acceptance Criteria Certified Design Commitment Inspections. Tests. Analyses Acceptance Criteria

1. He flow paths required to provide safety Inspections shall be cenducted of the as-built De as-built configuration of the Dow paths for injection are in-laded in the as-built accumulator tank subsystem. the accumulator subsystem is as follows:

configuration for the accumulator tank subsystem. a. An outlet discharre line connects to ene of the direct vessel injection lines which injects into the reactor vessel downeomer.

b. Each accumulator dir. charge line contains a dic. charge isolatien valve. two series check valves. and a flow tuning orifice.

2. The accumulator discharge isolation vahes Functional tests shall be conducted ne discharge isolation s alves open upon cpen upon receipt of an actuation signal from demonstrating proper valve operation upon receipt of an actuation signal.

the protection and safety monitoring system. receipt of an actuation signal from the protection and safety monitoring system.

3. Le division assignment for valves and With only the assigned Class IE power and ne accumulator discharge line isolation valve centrols is such that the loss of any single protection division energized, tests shall be assigned to the energized division receive an Class IE division will not prevent the conducted to confirm power and protection actuation signal and have power available.

accumulator tank subsystem from perform.ing division assignments by operating the control its safety-related function of providing safety circuits. His test shall be performed for each injection. assigned power division.

W Westinghouse

o J ( U Tier 1 Derign Certification Material PASSIVE CORE COOLING SYSTEM ~

Revision: 1 _

Effective: 12/15/92 t

I Table 3.2.4-1.3 - Passive Core Cooiing System (Accumulators)

Inspections, Tests, Analyses and Acceptance Criteris - _=

inspections, Tests, Analyses Acceptance Criteria , ,

Certified Design Commitment j e

A low pressure injection test for each De accumulator discharge flow rate from each

4. Each accumulator provides the regtured accumulator shall be conducted to determine the accumulator at a piessure of [200J psig is:

injection flow to the reactor coolant system.

accumulator injection flow capability.

2 {4300] gpm Me reactor vessel head and internals shall be s; [4900] gpm removed and the ra .:. vuoiant system shall be empty.

Each accumulator shall be filled with water to between {l000] and [1100] ft' and pressurized with nitrogen to between [270] and [300] psig.

Tests or as-built dimensions and volumetric The volume of each accumulator is

5. He accumulator volume shall be sufficient analyses shall be performed to verify the tara 2 [2000] cubic feet.

to provide the required injection.

volume.

l i

I 3.2.4-14 W ,Westinghouse

m G

Tier 1 Design Certification Material PASSIVE CORE COOLING SYSTEM 1.

E Revision: 1

=

Effective: 12/15/92 ..

Table 3.2.4-1.4 - Passive Core Cooling System (IRWST and Containment Recirculation)

Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections Tests, Analyses Acceptance Criteria

1. He flow paths required to provide safety Inspections shall be condorred of the as-built a. De as-built configuration of the flow paths injection are included in the as-built IRWST and containment recirculation for the IRWST and containment recirculation configuration for the IRWST and subsystem. subsystem is as follows:

containment rxircu'ation subsystem.

(1) Two injection lines from the IRWST.

ene to each of the two direct sessel injection lines which inject into the i reactor vessel dou n:omer. Each injection line contains an isolation valve. a flow turing orifice, and four check valves (two pasallel paths of two series check valves).

I (2) Each IRWST injection line connects to an associated recirculation screen.

De connection, which is upstream of the IRWST injecticn line isolation valve, consists of twc ' ~2 M recirculation lines. On  : tw o parallel recirculation !W .;om each screen contains two series check valves and the other line contains two series isolation valves. 1 W-Westingflouse

p N O(~

Tier 1 Design Certification Material

": =

PASSIVE CORE COOLif4G SYSTEM

• Revision: 1 Effective: 12/15/92 ,

Table 3.2.4-1.4 - Passive Core Cooling System (IRWST and Containment Recirculation) inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests, Analyses Acceptance Criteria l 1

1. (continued) b. He elevation of the overflow level in the y IRWST is higher than the direct vessel injection nozzle centerline by:

2 {32] feet

2. Ac IRWST injection line discharge isolation t. Functional tests shall be conducted a. The IRWST injection line discharge isolation valves open upon receipt of an actuation demonstrating proper valve operation in the valves open upon receipt of an actuation signal from the protection and safety presence of an actuation signal from the signal.

monitoring system and the diverse actuation protection a d safety monitoring system.

system.

b. Functional tests shall be conducted b. He IRWST injection line discharge isolation demonstrating proper valve operation in the valves open upon receipt of an actuation presence of an actuation signal from the signal.

di"erse actuation system.

3. The containment recirculation line isolation Functional tests shall be conducted he containment recirculation line isolation valves open upon receipt of an actuation demonstrating proper valve operation in the valves open upon receipt of an actuation signal.

signal from the protection and safety presence of an actuation signal from the monitoring system. protection and safety monitoring system.

W Westinghouse

ry Tier 1 Design Certification Material PASSIVE CORE COOLING SYSTEM Revision: 1 ,

l Effective: 12/15/92 Table 3.2.4-1.4 - Passive Core Cooling System (IRWST and Containment Recirculation)

Inspections, Tests, Analyses and Acceptance Criteria p inspections, Tests, Analyses Acceptance Criteria Certified Design Commitment l

with only the assigned Class IE power and ne IRWST injection line isolation valve and l

4. He division assignment for valves and protection division energized, tests shall be the containment recirculation line isolation controls is such that the loss of any single l

conducted to confirm power and protection valve assigned to the energized division receive Class 1E division will not prevent the l

division assignments by operating the control rn actuation signal and have power available.

IRWST and containment recirculation subsystem from performing its safety-related circuits. This test shall be performed for each function of providing safety injection. aasigned power division.

A test for each IRWST injection line shall be ne IRWST flow rate from each injection line

5. De IRWST provides the required injection into the direct vessel injection line connection, flow to the reactor coolant system. conducted to determine the injection flow capability. with the IRWSl~ water level between [15.90]

and [16.10] feet above the direct vessel The reactor head and internals shall be removed injection nozzle centeiline, is:

and the reactor coolant system shall be empty.

He IRWST shall be partially filled with water. 2 [345] gpm with a water level at least [18] feet above the s;; [1000] gpm direct vessel injection nozzle centerline.

W

- Westin=ehouse

_ _ _ _ l

Tier 1 Design Certification Material PASSIVE CORE COOLING SYSTEM Revision: 1 _

Effective: 12/15/92 Table 3.2.4-1.4 - Passive Core Cooling System (IRWST and Containment Recirculation)

Inspections, Tests, Analyses and Acceptance Criteria inspections, Tests, Analyses Acceptance Criteria Certified Design Commitment A test for each containment recirculation screen The containment recirculation flow rate fer

6. The containment recirculation lines provide line shall be conducted to determine the flow each recirculation !ine into the direct vessel the required recirculation flow to the reactor injection line connectien, with the injection coolant system. capability.

water leve! betw een [7.90] and [S.10] feet above the direct vessel injection nozzle

'Ihe reactor head shall be removed and the reactor coolant system shall be empty. A centerline, is:

temporary water supply shall be connected to the recirculation line with a water level of at 2 [t00] gpm least [10] feet above the direct vessel injection s [455] ppm nonle centerline.

Tests or as-built dimensions and solumetric The volume of the IRWST below the tank

7. 'Ihe IRWST volume is sufficient to provide analyses shall be performed to verify the overDow elevation is 2 [7N00] cubic feet.

the required injection.

IRWST volume.

Tests or as-built dimensions and volumetric %e total floodable volume in the centainment S. He containment flooded volume and for all compartments, with the exception of the elevation head sha!! be sufficient to provide analyses shall be performed to verify the chemical and volume control system equipment the required recirculation. floodable containment volume.

areas, at an elevation that is [8] feet above the direct vessel injection nonle centerline is:

s [TBD] cubic feet W-Westinehouse

=

Tier 1 Design Certification Material PASSIVE CORE COOLING SYSTEM Revision: 1 _

Effective: 12/15/92 Table 3.2.4-1.5 - Passive Core Cooling System (pH Adjustment Tank)

Inspections, Tests, Analyses and Acceptance Criteria inspections, Tests, Analyses Acceptance Criteria Certified Design Commitment Inspections shall be conducted of the as-built pH

a. De as-built configuration of the flow paths

1. He flow paths required to provide for the pH adjustment tank subsystem is as containment pli control are included in the adjustment tank subsystem.

follows:

as built configuration for the pil adjustment tank subsystem.

(1) He tank inlet line cor.tains two parallel vacuum breakers.

(2) A discharge line connects to the tank bottom and drains to both containment recirculation screen areas.

De pH adjustment tank discharge flow path contains two paral!el isolation valves.

He discharge flow path divides into two cross <onnected lines, downstream of the isolation valves, to provide flow to bott containment screen areas.

b. De tank bottom (outside surface) is Io ated 2 [12] feet above the direct vessel injection nozzle centerline.

W -

Westinghouse l y

~

f V v Tier 1 Design Certification Material v

PASSIVE CORE COOLING SYSTEM Revision: 1 Effective: 12/15/92 _

_y Table 3.2.4-1.5 - Passive Core Cooling System (pH Adjustment Tank)

Inspections. Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

2. The pl1 adjustment tank vacuum breakers Functional tests shall be conductM a. He vacuum breaker opens and the discharge self-actuate opca and the discharge isolation demonstrating valve operation in the presence of valve opens, as confirmed by observing pH valves open upon receipt of an actuation an actuation signal. adjustment tank flow through the discharge signal from the protection and safety lines to both containment recirculation screen monitoring system. The pli adjustment tank shall be partially filled areas.

with water, at atmospheric pressure, without nitrogen makeup to the tank. b. The pH adjustment tank emptin in s [100] minutes.

A separate test shall be conducted for each discharge isolation valve. One vacuum breaker c. De volume of water that drains to each valve shall be gagged for each test. screen area is 2 [40] percent of the initial tank test volume.

3. He division assignment for valves and With only the assigned Class !E power and ne discharge line isolation valve assigned to controls is such that the loss of any single protection division energized, tests shall be the energized division receive an actuation Class 1E division will not prevent the pli conducted to confirm power and protection signal and have power available.

adjustment tank from performing its safety- division assignments by operating the control related function of providing containment pli circuits. His test shall be performed for each control. assigned power division.

4. He pl{ adjustment tank volume is sufficient Tests or as-built dimensions and volumetric ne total pli adjustment tank volume is to provide the required containment pH. analyses shall be performed to verify the pl{ 2 [160] cubic feet.

adjustment tank volume.

W-Westinghouse

Tirr 1 Dnign C:rtific: tion M:.terial STEAM GENERATOR SYSTEM (V) Revision: 1

~

F Effutive: 12/15/92 3.2.5 STEAM GENERATOR SYSTEM Design Description The Steam Generator System (SGS) is a The steam generator system performs the following safety-related system that is made up of the safety-related and defense-in-depth functions, safety-related portions of the main steam, main feedwater, startup feedwater and blowdown lines Safety-related functions:

connected to the steam generator. The safety-related portions of the steam generator system include:

• The steam generator system provides overpressure protection for the steam generator secondary side,

• That portion of each main steam line from the as well as portions of the main steam line, steam generator outlet nonle up to and including feedwater lines and blowdown lines by means of the the piping anchors at the auxiliary building / turbine safety relief valves for events that result in steam building wall. Within this safety-related portion of generator pressure transients. Three safety valves the main steam line are the main steam and bypass are provided on each main steam line, isolation valves and connected upstream of these valves, the power operated relief and block valves,

• During design basis accidents, the steam generator the condensate drain isolation valves and the three system in conjunction with the main steam system j [ safety valves. prevents excessive steam generator blowdown and i ( excessive feedwater flow from the main and startup l

• That portion of each main feedwater line from the feedwater system. This function is accomplished steam generator feedwater nonle up 9 and via isolation of the main feedwater, startup including the feedwater control valve. Within this feedwater, blowdown and main steam lines, portion of the feedwater line are the main feedwater isolation valve and the downstream feedwater check

• In the event of feedwater unavailability, the steam valve. generator system removes decay heat from the reactor coolant by releasing the steam generated

• That portion of each startup feedwater line from the from the steam generator inventory to the connection to feedwater line out to the restraint at atmosphere via the safety valves, the auxiliary building / turbine building wall. The line includes the startup feedwater control valve and

• The portion of the steam generator system inside a downstream startup feedwater isolation valve. containment is an integral part of the containment isolation boundary and limits radioactive releases to

• That portion of each steam generator blow down line the environment. The inside containment isolation from the steam generator blowdown outlet nonle function (isolating the reactor coolant system and up and including the second steam generator containment atmosphere from the environment) is isolation valve. Within this safety-related portion of provided by the steam generator, tubes, and steam the line are the two remotely operated isolation generator system lines inside containment, while valves in series. isolation outside containment is provided by manual and automatic valves. The containment isolation

/m\ 3.2.5-1

!v/ W Westinghouse

Ti:r 1 Dnign C:rtification M:t: rial STEAM GENERATOR SYSTEM

+'  ! Hit Revision: 1 Effective: 12/15/92 safety-related function is covered in the containment system Tier 1 Design Description in Section 3.2.2.

• %e steam generator system, in conjunction with passive core cooling and chemical and volume control system safety features is designed to avoid steam line flooding after a steam generator tube j rupture accident and thus serves to limit the releases to the environment. The steam generator system features include provisions for feedwater and startup feedwater isolation.

Defense-in-Depth:

• During shutdown operations, the steam generator system removes decay heat by delivery of feedwater from the startup feedwater system to the steam generator and venting of steam from the steam generators to the atmosphere via the power operated relief valves, ~j 1

The valves and cont ols necessary for system actuation and control are powered from Class 1E de power The divisiou usignment for valves and controls is such that the loss of any single Class IE power ,

division will not prevent system safety functions.

%e steam generator system in combination with other safety-related systems is designed such that no ,.;

single failure in the steam generator system will prevent the systems from accomplishing identified safety functions, In order to provide accident mitigation functions, the steam generator system valves close on the appropriate isolation signal. He isolation signal may be generated automatically or manually at a system level and manually at the valve level within the protection and safety monitoring system. He manual protection and safety monitoring system signals can be initiated from either the main control room or the remote shutdown workstation.

3.2.5-2 W

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v Tier 1 Design Certification Material STEAM GEtJERATOR SYSTEM 7 -

Revision: 1 Effective: 12/15/92 _

Table 3.2.5 Steam Generator System inspections. Tests, Analyses and Acceptance Criteria -

Certified Design Commitment inspections, Tests. Analysis Acceptance Criteria

1. He SGS provides overpressure protection a. Inspections shall be conducted to confirm a. ne sum of the rated capacities of the t nee for the secondary sitie of the steam valve vendor code plate rating is greater valves in each main steam line exceeds [4.2 generator and the SGS piping by means of than or equal to system relief requirements. x 10'] lbzhr as recorded on the valve the safety relief valves. vendor code plates.

b. Tests shall be conducted to confirm the b. He set pressures of the three safety valves safety valve provides overpressure are:

protection.

[10S5] psig - [20] psig

[1113] psig i [20] psig

[1140] psig - [20] psig

2. %e SGS isolates the main feedwater, Valve functional tests shall be conducted He main steam isolation, feedwater isolation startup feedwater, blowdown and main demonstrating proper valve operation in the and feedwater control valves close within steam lines. He isola: ion valves close presence of an actuation signal. [5] seconds upon receipt of an actuation signal.

upon receipt of manual or automatic actuation signals from the protection r.nd De startup feedwater isolation and control safety monitoring system. valves, blowdown isolation valves, MSIV bypass valve and condensate drain valves close within [10] seconds t>pon receipt of an actuation signal.

De power operated relief valve and PORV block valves close upon receipt of an actuation signal.

W_

Westinghouse

O O Tier 1 Design Certification Material O

STEAM GENERATOR SYSTEM =

Revision: 1 Effective: 12/15/92 .

Table 3.2.5 Steam Generator System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analysis Acceptance Criteria

3. He division assigmnent for isolation valves With only the assigned Class IE power and ne isolation valves within the main steam and controls is such that the loss of any protection division energized, a test shall be lines, main and startup feedwater lines, and single Class !E division does not prevent conducted to confirm power and protection blowdown lines assigned to the energized system safety function. division assignments by operating the control division receise power and signals.

circuit. His test shall be performed for each assigned power division.

4. He SGS removes decay heat by delivery of Valve functional tests shall be conducted He startup feedwater isolation valve and feedwater from the startup feedwater demonstrating proper valve operaticn in the control valve and the power operated relief system to the steam generator and steam rr=ence of actuation signals from the plant valve and block valves open upon receipt of an from the steam generators to the control system or the protection and safety actuation signal.

atmosphere. monitoring system.

.2.54 W Westinghouse

Tirr 1 D: sign C rtific:: tion M;t rial lg) MAIN CONTROL ROOM EMERGENCY HABITABILITY SYSTEM

\._) Revision: 1 5' Effectivo: 12/15/92 3.2.6 MAIN CONTROL ROOM EMERGENCY HABITABILITY SYSTEM Design Description he main control room emergency habitability conjunction with the automatic closing of the system (VES) is a safety-related system that provides nuclear island nonradioactive ventilation system breathable air and pressurization for the main control main control room isolation dampers. Refer to the room. The VES also provides passive heat sinks for the nuclear island nonradioactive ventilation system main control room, the protection and safety monitoring Tier i Design Description in Section 3.7.1, Re system instrumentation and control rooms, and the same refilling connection used for providing Class IE de equipment rooms. It consists of two breathable air is provided to allow for operation redundant trains, each with the following: longer than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

• Compressed air storage

• Re main control room emergency habitability system provides passive heat sinks for the main

• A pressure contml valve and a flow control ontice control room, the protection and safety monitoring located in the supply pipe connecting the system instrumentation and control rooms, and the compressed air storage to the main control room Class IE de equipment rooms. The heat sinks for envelope these rooms consist of the surrounding concrete

[b \

• A remotely operated isolation valve located downstream of the flow control orifice.

structure and provide cooling capacity for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The design of these rooms allows for the use of a temporary cooling source for the post-72 hour period.

When the nuclear island nonradioactive ventilation system is not available, the main control room Re valves necessary for system actuation are emergency habitability system performs the following powered from the Class IE de and UPS system. The safety-related functions: division assignment for valves is such that the loss of any single Class IE power division will not prevent the

• The main control room emergency habitability system from accomplishing its safety-related functions.

system provides breathable air for the occupants of No single failure will prevent the main control room the main control room. This function is emergency habitability system from accomplishing the accomplished via automatic opening of either identified safety-related functions.

isolation valve. The initial capacity of air in the In order to provide accident mitigation functions, compressed air storage tanks provides for at least the main control room emergency habitability system 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of operation. A safety-related refilling valves open on the appropriate actuation signal. The connection is provided to allow for operation longer actuation signal may be generated automatically at the than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. system level and manually at the valve level within the protection and safety monitoring system. The manual

• The main control room emergency habitability signal can be initiated from the main control room, system provides pressurization of the nuin control room envelope. This function is accomplished via automatic opening of cither isolation valve in CT 3.2E-1 iv / W-Westinghouse

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Tier 1 Design Certification Material J

MAIN CONTROL ROOM EMERGENCY HABITABILITY SYSTEM _

Revision: 1 Effective: 12/15/92 .

Table 3.2.6 Main Control Room Emergency Habitability System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections Tests, Analyses Acceptance Criteria

1. De VES provides breathable air for the a. Tests shall be conducted to confirm that the a. He air flow rate from one train of the VES occupants of the main control room. combined action of the pressure control is greater than or equal to [20] scfm and valve and flow control orifice deliver the less than or equal to [22] scfm.

required amount of air flow to the main control room.

b. Tests shall be conducted to confirm that the b. He air flow rate from the temporary source required amount of air flow can be delivend of air is greater than or equal to [20] scfm.

to the main control room from a temporary source via the refilling connection.

2. He VES provides pressurization of the Tests shall be conducted to confirm that the air De main control room envelope is pressurized main control room er velope. flow supplied by the VES produces the required to greater than or equal to [l/8] inch water pressurization of the main control room gage with respect to the surrounding area.

envelope.

3. The VES provides passive heat sinks for the Using as-built information and design basis heat the temperature rise for the main control room main control room, the protection and safety loads, a heat sink capacity analysis shall be is less than or equal to [20]* F for the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> monitoring system instrumentation and performed for each room. period. De maximum temperature for the control rooms, and the Class 1E de instrumentation and control rooms and the de equipment rooms. equipment rooms is less than or equal to

[120]* F for the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period.

4. De VES isolation valves open upon receipt A functional test shall be conducted on each Each VES isolation valve opens.

of the specified actuation signals from the . isolation valve using a manual protection and protection and safety monitoring system. l safety monitoring system actuation signal.

.2 + 2 w westingnouse

%j Y Tier 1 Design Certification Material MAIN CONTROL ROOM EMERGENCY HAB!TABILITY SYSTEM _

Revision: 1 Effective: 12/15/92 Table 3.2.6 Main Control Room Emergency Habitability System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

5. He division assignment f >r the isolation With only the assigned Class IE power and The VES isolation valves assigned to the valves is such that the loss of any single protection division energized, tests shall be energized division receive an actuation signal Class 1E division will not prevent the VES conducted to confirm power and protection and have power available.

from accomphshing its safety-related division assignments by operating the control functions. circuits. This test shall be performed fer each assigned power division.

6. Compressed air storage capacity provides An inspection and analysis of the as-built At lent io6400] sef is provided.

breathable air and pressurization of the main compressed air storage capacity shall be y u,a:mt room envelope for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. performed.

.2.6-3 W-Westinehouse u

Tier 1 Design Certification Meterial COMPONENT COOLING WATER SYSTEM Revision: 1 n:- '@

~

T Effective: 12/15/92 3.3.1 COMPONENT COOLING WATER SYSTEM Design Description The component cooling water system (CCS) serves the component cooling water pumps, heat exchangers, no safety-related function except for containment surge tank, and the piping associated with the normal isolation. The containment isolation safety-related residual heat removal heat exchangers and pumps, spent function is covered in the containment system Tier 1 fuel pit heat exchangers, and makeup pump miniflow Design Description in Section 3.2.2. heat exchangers.

'ne component cooling water system performs the ,

following defense-in-depth functions:

• Provides cooling for the normal residual heat z removal heat exchangers and pumps for the removal of heat from the reactor coolant system during plant cooldown and refueling operations when the reactor coolant system pressure and temperature are below 450 psig and 350*F.

• Provides cooling for the rniniflow heat exchangers of the chemical and volume control system makeup pumps.

• Provides cooling for the spent fuel pit heat exchangers for heat ternoval from the spent fuel pit.

The componer t cooling water system is a closed loop system that transfers heat from various nonsafety-related plant components to the service water system.

The component cooling water system is powered from luth onsite and offsite ac sources. The system consists of redundant component cooling water pumps and heat exchangers, and associated valves, piping, and instrumentation. A surge tank is connected to the return header.

The components of the component cooling water system which provide the defense-in-depth functions are 3^

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p (\ ,y b

Tier 1 Design Certification Material COMPONENT COOLING WATER SYSTEM  %' :g

~

Revision: 1 Effective: 12/15/92 _

i Table 3.3.1 Component Cooling Water System inspections. Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. He component cooling water system A component cooling water flow test shall be He flow rates to the components are greater provides flow to the normal residual heat performed to verify cooling water flow rates to than or equal to:

removal pumps and hat exchangers, the the components supporting defense-in-depth spent fuel pit heat exchangers, and the functions. These components are the normal Normal Residual Heat Removal Heat Exchanger makeup pump miniflow heat exchangers. residual heat removal heat exchangers and [16(X)] gpm pumps, the spent fuel pit heat exchangers, and the makeup pump miniflow heat exchangers. Normal Residual Heat Removal Pump

[4] gpm He test shall be conducted with one component cooling water pump operating and flow through Spent Fuel Pit Heat Exchanger one normal residual heat removal heat exchanger [400] gpm and pump, one spent fuel pit heat exchanger, and one miniflow heat exchanger. He test shall Makeup Pump Miniflow Heat Exchanger be repeated with the other component cooling [30] gpm water pump.

Westirighouse

O A

%J '

Tier 1 Design Certification Material COMPONENT COOLING WATER SYSTEM Revision: 1 Effective: 12/15/92 .

Table 3.3.1 Component Cooling Water System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

2. The component cooling water system A test and associated analyses shall be ne calculated component cooling water heat provides cooling for the normal performed to verify the heat removal capability exchanger minimum heat rejection capability residual heat removal pumps and heat of a single component cooling water heat resulting from the analyses is at least exchar.gers, the spent fuel pit heat exchanger to perform defense-in-depth functiens [100 X 10+] Btu /hr.

exchangers, and the makeup pump at site-limiting environmental conditions. The miniflow heat exchangers. test can be conducted at any site conditions and the results analyzed to confirm heat exchanger prformance at the limiting site conditions. His test shall be repeated with the other component cooling water heat exchanger.

3. He component cooling water surge An inspection of the as-built component cooling He elevation of the outside bottom surface of tank location provides adecluate net water system shall be conducted. the surge tank is at least [30] feet above the l positive suction head for the component centerline of the component coolinf water pump I ecoling water pumps. . suction nozzles. l T Westinghouse

Tiir 1 D: sign Certific: tion Mitirial n

[ \ CHEMICAL AND VOLUME CONTROL SYSTEM

\. Revision: 1 .

Effective: 12/15/92 3.3.2 CHEMICAL AND VOLUME CONTROL SYSTEM Design Description The chemical and volume control system (CVS) is Reactor Coolant System Isolation a nonsafety-related system except for the portions that perform the following safety-related functions: The cheadcal and volume control system has redundant, safety-related isolation valves and piping to a preservation of the reactor coolant system pressure protect the reactor coolant system pressure boundary, boundary, including isolation of normal chemical %ese valves are part of the reactor coolant pressure and volume control system letdown from the reactor boundary. Letdown stop valves in the line to the coolant system reactor coolant system cold leg are remotely operated valves. Rese valves close upon receipt of an actuation

• Containment isolation of chemical and volume signal from the protection and safety monitoring system control system lines penetrating containment. The or manual control from the main control room. The containment isolation safety-related function is makeup line from the regenerative heat exchanger to the covered in the containment system Tier i Design reactor coolant system contains a check valve and a Description in Section 3.2.2. remotely-operated stop check valve for isolation. The line from the chemical and volume control system to the g\

• Termination of inadvertent reactor coolant system pressurizer auxiliary spray has a check valve and a l

(O boron dilution remotely-operated stop check valve for isolation. Ee stop check valve is operated from the main control

• Isolation of makeup. room.

He chemical and volume control system provides Boron Dilution Termination the following defense-in-depth functions:

The chemical and volume control system terminates

• Supply makeup to the reactor coolant system. a boron dilution accident by closing redundant, safety-related, remotelymperated valves in the line from the

• Supply coolant to the pressurizer auxiliary spray demineralized water system to the makeup pumps and line, aligning the three-way pump suction control valve such that the makeup pumps take suction from the boric acid he portions of the system that provide tank upon receipt of a signal from the protection and defense-in-depth functions include the makeup pumps safety monitoring system.

and the boric acid tank that supplies the pumps. The makeup pumps supply water to the reactor coolant Mokeep Isolation system loop and the pressurizer spray. The pumps are powered from both onsite and offsite ac power sources. To protect against steam generator overfill Connecting piping and selected isolation and stop valves following a postulated steam generator tube rupture and are required for safety-related and defense-in-depth pressurizer overfill, the makeup function is isolated by functions. closing the makeup line containment isolation valves O 3.3.2-1 W Westinghouse iv) -

Tier 1 D: sign CIrtification M:,t: rial CHEMICAL AND VOLUME CONTROL SYSTEM .

$ Revision: 1 Effective: 12/15/92 when an appropriate signal is generated by the protection and safety monitoring system.

Defense-in-Depth Functions

%e chemical and volume control system provides makeup. The chemical and volume control system makeup pumps are initiated upon receipt of a signal from the plant control system.

O 3.3.2-2 W Westinghouse G

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Tier 1 Design Certification Material

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CHEMICAL VOLUME AND CONTROL SYSTEM g

~

Revision: 1 Effective: 12/15/92 _

l Table 3.3.2 Chemical and Volume Control System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. De letdown stop valves close upon receipt Valve functional tests shall be conducted ne letdown stop vahes close upon receipt of of an actuation signa! from the protection demonstrating valve operation in the presence of an actuation signal.

and monitoring system. an actuation signal.

2. De auxiliary spray isolation valve Valve functional tests shall be conducted he auxiliary spray valves reposition upon repositions on receipt of an actuation signal demonstrating valve operation in the presence of receipt of an actuation signal.

from the main control room. an actuation signal.

3. He demineralized water system isolation Valve functional tests shall be conducted He valves close upon receipt of an actuation valves close upon receipt of en actuation demonstrating operation in the presence of an signal.

signal from the protection and safety actuation signal.

monitoring system.

4. He makeup pump suction header valve Valve functional tests shall be conducted He valve aligns pump suction to the boric acid repositions upon receipt of an actuation demonstrating valve operation in the presence of tank upon receipt of an actuation signal.

signal from the plant control system. an actuation signal.

5. The makcup pumps start and provide flow Pump functional tests shall be conducted Each pump starts upon receipt of and actuation upon receipt of an actuation signal from the demonstrating pump operation in the presence of signal and individually delivers at least plant control system. an actuation signal with reactor coolant system [100] gpm minimum flow through the makeup pressure of at least [TBD]. line with reactor coolant system pressure greater than [TBD]. l W -

Westinehouse =

%Y Tier 1 Design Certification Materini Y

CHEMICAL VOLUME AND CONTROL SYSTEM -

,g Revision: 1 Effective: 12/15/92 Table 3.3.2 Chemical and Volume Control System inspections Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, A7alyses Acceptance Criteria

6. The division assignment for the letdown With only the assigned Class IE power and %e chemical and volume control system valves isolation valves and controls is such that the protection division energizei a test shall be assigned to the energized division receive loss of any single Class 1E division does conducted to confirm power and protection power and signals.

not prevent letdown isolation. division assignments by operating the control circuit. His test shall be performed for each assigned power division.

7. He division assignment for the With only the assigned Class IE power and he chemical and volume control system valves demineralizer water isolation valves and protection division energized, a test shall be assigned to the energized division receive controls is such that the loss of any single conducted to confirm power and protection power and signals.

Class IE division does not prevent division assignments by operating the control demineralizer water isolation. circuit. His test shall be performed for each assigned power division.

'~

W-Westinghouse

Tier 1 Design Certification Material l

STANDBY DIESEL AND FUEL OIL SYSTEM Revision: 1 '.

Effective: 12/15/92 3.3.3 STANDBY DIESEL FUEL OIL SYSTEM Design Description The standby diesel fuel oil system (DOS) serves no safety-related function. The DOS supplies diesel fuel for the defense-in-depth function of the onsite standby power system.

The standby diesel fuel oil system is designed to store and transfer diesel fuel oil for the onsite standby power system. The system provides diesel fuel oil storage and a s:parate transfer flow path for each of the two diesel generators of the onsite standby power system. Each of the two subsystems includes a separate ,

supply connection from the storage tank (s) and a A transfer piping system containing a suction strainer, a transfer fuel cil pump, a fuel oil heater, a fuel oil moisture remover, a fuel oil filter and interconnecting piping discharging to its diesel generator day tank in the onsite standby power system. The transfer fuel oil pump and the fuel oil heater can be powered from both onsite and offsite ac sources. The storage tank provides reserved fuel oil storage for each diesel for at least seven days of continuous diesel generator operation.

W Westinghouse

g

'N N.,) Q Tier 1 Design Certification Material

=H STANDBY DIESEL AND FUEL Olt SYSTEM n Revision: 1 .

Effective: 12/15/92 Table 3.3.3 Standby Die.sel Fuel Oil System

!nspections, Tests, Analyses and Acceptance Criteria inspections. Tests, Analyses Acceptance Criteria Certified Design Commitment

]

he DOS tank storage reserved for each Diesel ne adequacy of the DOS tank storage shall be '

1. De DOS tank (s) store (s) fuel oil for Generator is at least [34,500] gallons.

seven days of operation of each diesel validated by as built measurement and associated analysis of the stored fuel volume from abose generator.

the tank bottom freeboard to the top level of fuel storage reserved for the diesel generators.

ne flow delivered to the diesel generator day De DOS provides fuel oil flow to each of The DOS perfonnance shall be validated by a

2. tank is 2 [18} gpm.

functional test. He DOS fuel delivery rate shall the standby diesel generator day tank to be determined by the measured time required to support diesel generator continuous fill the diesel generator day tank between two operation at the load capacity identified in the onsite standby power system measured levels.

Inspections, Tests, Analyses and Acceptance Criteria Table 3.6.5-1, entry number 3.

3. 3-2 W Westinghouse

Ti;r 1 D: sign Ccrtificnion Mit: rial j FIRE PROTECTION SYSTEM _.

V Revision: 1 E -

Effective: 12/15/92 e

3.3.4 FIRE PROTECTION SYSTEM Design Description he fire protection system (FPS) is a nonsafety-related system and it has no safety-related function other than containment isolation. The containment isolation safety-related function is covered in the containment system Tier i Design Description in Section 3.2.2.

The fire protection system performs the defense-in-depth function of providing equipment for manual firefighting in areas containing equipment required for safe shutdown.

The fire protection standpipes which support the defense-in-depth function of the fire protection system are supplied with water by gravity from the passive containment cooling system storage tank and function independently of the rest of the fire protection system.

Refer to the passive containment cooling system Tier 1

[]

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Design Description in Section 3.2.3. He supply line draws water from the upper portion of the storage tank, using water allocated for fire protection. The system does not require electrical power since it relies on gravity and is numually actuated.

(~N! W Westinghouse 3.3.4-1 iv -

7-V V U Tier 1 Design Certification Material FIRE PROTECTION SYSTEM Revision: 1 Effective: 12/15/92 Table 3.3.4 Fire Protection System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. The supply line draws water from the Each standpipe supplied from the passive Water is dixharged from each of the two hoses passive containment cooling system storage containment cooling system storage tank shall be at not less than [75] gallons per minute.

tank and delivers water through the individually tested. W ater shall be standpipes. imultaneously discharged from the two highest hose stations on the standpipe. The normally installed hoser and nozzles shall be used.

Adjustable nozzles may be adjusted to demonstrate that required flow can be achieved.

3. . 2 W Westinghouse

Ti:r 1 D: sign C:rtific tion M;terial

,o MECHANICAL HANDLING SYSTEM (k -}

-- Revision: 1 .

Effective: 12/15/02 3.3.5 MECHANICAL HANDLING SYSTEM Design Description The mechanical handling system (MHS) components that are safety-related are the containment polar crano and the spent fuel shipping cask crane. The safety function of these cranes is to prever t uncontrolled lowering of heavy loads in safety-reisled areas. A heavy load is a load whose weight is greater than the combined weight of a single spent fuel Assembly and its handling device.

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G G Tier 1 Design Certification Material v

MECHANICA1. HANDLING SYSTEM -

Revision: 1 Effective: 12/15!92 _

Table 3.3.5 Mechanical Handling Systern inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria F 1. The containment polar crane pres ents a. The upper travel limit switches for each of the a. Power is interrupted to the hoist motor and uncontrolled lowering of a heavy load, main and auxiliary hoists that handle heavy the ho;st holding brakes are set.

loads shall be tested individually. He hoist shall not be loaded for this testing.

b. He overspeed detection device (s) for each of b. Actuation of the overspeed detection the main and auxiliary hoists that handle heavy device (s) results in interruption of power to loads shall be tested. He hoist shall not be the hoist motor and setting of the hoist loaded for this testing. Special testing holding brakes.

methods based on the design of the detection device (s) may be used.

c. The hoist overload sensing system for each of c. De load sensing system causes power to the the main and auxiliary hoists that handle heavy hoist motor to be interrupted.

loads shall be tested. Special testing methods based on the design of the overload sensing system may be used.

(V Westinghouse

p s T U d Tier 1 Design Certification Material V

MECHANICAL HANDLING SYSTEM ii Revision: 1 "

Effective: 12/15/92 Table 3.3.5-1.- Mechanical Handling System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. (continued) d. He redundant hoist holding brakes for each of d. Each hoist holding brake stops and holds the the main and auxiliary hoists that handle heavy test load, loads shall be tested individually, with the hoist lowering a test load at the maximum operating speed for a heavy load. He test load shall not be less than the heaviest load to be handled by the hoist, but not more than the hoist load rating.

e. A functional test of the emergency load e. He test load is lowered to the floor with at lowering capability shall be performed for least [one] stop at an intermediate point each of the main and auxiliary hoists that during the lowering process.

handle heavy loads. De test load shall not be less than the heaviest load to be handled by the hoist, but not more than the hoist load rating.

He test load shall be lowered from a height of not less than [10] feet, or from the maximum height permitted by the configuration of the test load, whichever is less.

3' ' ~

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gm p s N,.

Tier 1 Design Certificttion Material MECHANICAL HANDLING SYSTb1 t=

I!i -

Revision: 1 Effective: 12/15/32 .

Table 3.3.5 Mechanical Hand!ing System Inspections, Tests, Analyses and Acceptance Criteria Certified Desiga Commitment insp:. .,Jons, Tests, Analyses Accep*wnce Criteria

1. (continued) f. The crane shall be load tested, using each of f. The crane iifts the test load, tramports it the main and auxiliary hoists that handle heavy through the required range of mo.mt.

Ioads. The test load shall not be more than lowers. stegw. and holds the load uith the

[125] percent of the hoist load rating. The hoi < holding brakes, required range of bridge and trolley movements shall, as a minimum. be representative of those required for the heavy loads to be handled by the hoist.

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Tier 1 Design Certifiution Materia.

g MECHANICAL. HANDLING SYSTEM U Revision: 1 Effective: 12/15/92 Table 3.3.5 Mechanical Handling Systern Inspections Tests, Analyses and Acceptance Criteria i

inspections. Tests. Analyses Acceptance Criteria Certified Design Commitment _=

He upper trave! limit switch-s for each of the a. Power is interrupted to the hoist meter and

2. He spent fuel shipping cask crane prevents a.

main and autiliary hoists that handle heavy the hoist ho! Jing brakes are set.

uncontrolled lowering of a heavy load.

loads shall be tested individually. De heist l

shall not be loaded for this testing.

l

b. He overspeeo detection device (s) for each of

b. Actuation of the os erspeed detec+ ion l the main and auxiliary hoists that handle heavy desicets) results in interruption of power to the hoist meter and setting of the hoist loads shall be tested. He hoist shall not be loaded for this test:ag. Special testing holding brakes.

methods based on the design of the detection device (s) may be used.

. The lotd wnsing =ystem caures power to the

c. We hoist overload sensing system for each of the main and auxiliary hoists that handle heavy hoist wtor to be interrupted.

loads shall be tested. Special testing rnethods based on the design of the overload sensing system may be used.

d. De redundant hoist holding brakes for each of d. Each hoist holding brake stqs and holds the the main and auxiliary hoists that handle heavy test lead.

I leads shall be tested individually, with the hoist lowering a test load at the maximum operating speed for a heavy load. De test W Westinghouse

O 4 Tier 1 Design Certification Material MECHANICAL HANDLING SYSTEM Revision: 1 Effective: 12/15/92 Table 3.3.5 Mechanical Handling System inspections. Tests, Analyses and Acceptance Criteria i

Certified Design Commitment I inspections, Tests, Analyses Acceptance Criteria l

2. teontinued) e. A functional test of the emergency load c. The test load is sowered to the floer with at lowering capability sha!! be performed for least [one) stcy at an intermediste point each of the main and auxiliary hoists that during tiie lowering process.

handle heavy loads. De test load shall not be less than the heaviest load to be handled by the heist, but not more than the hoist load rating.

He test load shall be lowered from a height of not less than [10] feet, or from the maximum height permitted by the configuration of the test load, whichever is less,

f. The crane shall be load tested. using each of f. The crane lifts the test load, transports it the main and auxiliary hoists that handle heavy through the required range of movement.

loads. De test load shall not be more than lowers, stop=. and holds the load with the

[125] percent of the hoist load rathg. He hoist holding brakes.

required range of bridge and trolley movements shall. as a minimum. be reprentative of those required for the heavy loads to be handled I;y the hoist.

3.3.5-6 3 Westinghottse

Ti:r 1 D: sign Ccrtific: tion M:terial b~

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PRIMARY R8 MPLING SYSTEM Revision Effectiv' 'M 192 3.3.6 PRIMARY SAMPLING SYSTEM Design Description 1he primary sampling system (PSS) is a non safety-related system except for containment isolation. 1he containment isolation safety-related function is covered in the containment system Tier 1 Design Description in Section 3.2.2. The primary sampling system is a defense-in-depth system which collects and delivers reprearntative samples.

1he PSS performs the following defense in-depth functions:

• The PSS obtains samples from th9 reactor coolant system loops and the containment recirculation inlet.

O V

1he defense-in depth portions of the primary sampling sy6 tem are:

• The sampling lines from the reactor coolant system loops and the containment recirculation inlet to the grab sampling unit

• The grab sampling unit up to and including interfaces with support auxiliaries

• 'Ihe sampling return lines from the grab sample unit to the containment recirculation area.

An actuation signal may be generated manually at the valve level within the protection and safety monitoring system to align the primary sampling system for sampling the reactor coolant system loops and the containment recirculation inlet.

3.3.G 1

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O Tier 1 Design Certification Material O

PRIMARY SAMPLING SYSTEM i!E Revision: 1 Effective:. 12/15/92 Table 3.3.6 Primary Sampling System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests. Analyses Acceptance Criteria

1. He PSS samples the reactor coolant system a. A test shall be conducted to confirm the a. De sampling system provides a flow rate locps and the containment recirculation senpling system's ability to obtain a sample of at least [0.51 gym.

inlet.  :!-au the reador coolant loops at I

atmosph ric i [0.2] psi pressure.

I o. With a water supply at the containment b. He sampling system provides a liow rate recirculation inlet a test shall be conducted of at least [0.5] cpm.

to confirm the sampling system's ability tc obtain a sample at atmospheric i [0.2} psi pressure.

2. He sampling system remotely operated Valve functional tests shall be conducted Each reactor coolant system hot Icg.

isolatien valves provide a sample flow path demonstrating proper val e operation upon containment recirculation inlet and contamment from the reactor coolant system loeps or receipt of an appropriate actuation sigr.al from isolation valve opens upon receipt of an containment recirculation inlet. the protection and safety monitoring system actuation signal and is subsequently clowd.

signal.

W Westinghouse

Ti:r 1 D: sign Certific: tion M:t:ri:1 i

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NORMAL RESIDUAL HEAT REMOVAL SYSTEM Revision: I E h

Effectivo: 12/15/92 4 a

3.3.7 NORMAL RESIDUAL HEAT REMOVAL SYSTEM Design Description 1he normal residual heat re:noval system (RNS) is residual heat removal pumps are powered from leth a nonsafety-related system except for those portions that onsite and offsite ac sources.

perfono the followint safety-related functions: The normal residual heat removal system has c~

redundant, safety-related isolation valves to protect the

• Prewrvation of reactar coolant system pressure reactor coolant system pressure boundary. These valves imundary integr.ty. are part of the reactor coolant pressure boundary.

There are two parallel sets of two valves in series

• Containment isolation of normal residual heat for a total of four valves for isolation in the line be-removal syste n lines penetrating containment. The tween the reactor coolant system hot leg and the residual containment isolation safety-related function is heat removal pumps. These valves are remotely-ce<ered in t',e containment system Tier i Design operated and are located inside containment. These Description in Section 3.2.2. valves are prevented from opening, by interlocks, whenever the reactor coolant system pressure exceeds lhe normal residual heat removal system provides the the normal residual heat removal system design following defense-in-depth functions: operating pressure. There are two check valves in series in each normal residual heat removal discharge Removes heat from the reactor coolant system line to each reactor coolant system direct vessel injection during operation at reduced pressure and line.

temperature. The normal residual heat removal system limits the in-containment refueling water storage tank water

• Provides low temperature overpressure protection for temperature during extended operation of the passive the reactor cooltnt system. residual heat removal system and limits the in contain-ment refueling water storage tank water temperature

• Provides cooling for the in-containment refueling during normal operation.

water storage tank. 1he nornal residual heat removal system provides low pressure malcup from the in-containment refueling

• Provides low prenure makeup to the reactor coolant water storage tank to the reactor coolant system. It system and removes heat from the reactor coolant provides reactor coolant system makeup once the pres-system following actuation of the automatic sure in the reactor coolant system is below the shutoff depressurization system. head of the normal residual heat removal system pumps.

During reactor coolant sysicm makeup operation, core The normal residu.1 heat removal system remoses decay heat removal is provided by aligning component both residual and sensible heat from the reactor coolant cooling water to the normal residual heat removal heat system. The normal residual heat removal system exchangers.

consists of residual heat removal pumps, residual heat removal heat exchangers, and the piping, valves and instrumentation necessary for system operation. The O

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Ti:r 1 D: sign C:rtific; tion M;terial NORMAL RESIDUAL HEAT REMOVAL SYSTEM Revision: 1

'F- Fffective: 12/15/92 The system can take suction f rom the reactor coolant system via a connection to one reactor coolant system hot leg and discharges to each reactor c(ulant system direct vessel injection line. 'The system can also take suction from the in-containment refueling water storage tank sia a single line and disebarges back to the in-containment refuehng water storage tank via a single line.

O 3.3.7-2 W Westinghouse O

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( 9 Tier 1 Design Certification Material NORMAL RESIDUAL HEAT REMOVAL SYSTEM Revision:.1 ij, Effective: 12/15/92 _

l Table 3.3.7 Normal Residual Heat Removal System  ;

Inspections. Tests. Analyses and Acceptance Criteria ,

Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. De reactor coolant system isolation valves Functional testing shall be performed to verify He RNS isolation valves do not open when '

are protected from inadvertently opening, by the valve cannot be opened when the reactor reactor coolant system pressure is greater than interlocks, wi.en the reactor coolant system _ coolant system pressure is above design [TBD] psig.

t pressure is above design operating pressure operating conditions of the RNS.

of the RNS.  ;

2. He RNS removes heat from the reactor Functional testing and associated analysis shall ne calculated heat transfer rate of the RNS i coolant system or the in-containment be performed to serify the heat removal capacity heat exchangers is at least ITBD] BTU /hr.

refueling water storage tank. of each RNS heat exchanger. He test can be [

conducted at any reactor coolant system j temperature between 100' F and 350" F and the l results converted to the limiting condition.

3. De RNS pumps take suctio from a reactor Functional testing shall be performed to verify He total system flow rate with both pumps coolant system hot leg and discharge to each the RNS flow rate with the pumps taking suction operating shall be greater than [TBD] gpm; the  !

reactor coolant system direct veswl injection from a reactor coolant system hot leg and total system flow rate with only one pump line. discharging to the reactor coolant system direct operating shall be greater than [TT>DJ gpm.

vessel injection lines. He test shall be performed with both pumps operating and with each pump operating individually. He test shall be perfonned with the reactor coolant system pressure less than [TBD] psig.

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Tier 1 Design Certification Material U

NORMAL RESIDUAL HEAT REMOVAL SYSTEM Revision: 1 Effective: 12/15/92 Table 3.3.7 Normal Residual Heat Removal System inspections. Tests. Analyses and Acceptance Criteria

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l Certified Design Commit nent inspections, Tests, Analyses Acceptance Criteria t

4. He RNS pumps take suction from the in- Functional testing shall be performed to verify De total system flow rate with lxwh pumps containment refueling water storage tank and the RNS flow rate with the pumps taking suction operating shall be greater than [TBD) gpm; the discharge to each reactor coolant system from the in<ontainment refueling water storage total system flow rate with only one pump direct vessel injection line. tank and discharging to the reactor coolant operating shall be greater than [TBD] gpm.

system direct vessel injection lines. He test thall be performed uith both pumps operating and with each pump operating individually.

5. He RNS pumps take suction from the in- Functional testing shall be performed to verify ne total system tiow rate with both pumps containment refueling water storage tank and the RNS flow rate with the pumps taking suction operating shall be greater than [TBD] gpm.

discharg- back to the in-containment from the in containment refueling water storage refueling water storage tank. tank and discharging back to the in-containment refueling water storage tank. He test shall be performed with both pumps operating.

6. He RNS relief valve provides low tempera- Test and analyses of the RNS relief valve ne rated discharge capacity of the RNS relief ture overpressure protection. capacity and set pressure shall be performed. valve is greater than [TBD] gpm at [TUD] psig.

W Westinghouse

Ti:t 1 D: sign Certific tion M:terial p) g SPENT FUEL PIT COOLING SYSTEM V Revision: 1 Effective: 12/15/92

_ t 3.3.8 SPENT FUEL PIT COOLING SYSTEM Design Description The spent fuel pit cooling system (SFS) is a nonsafety-related byttem except for those portions that perform the following safety-related functions.

• Containment isolation of the SFS lines penetrating the containment. The containment isolation safety related function is covered in the containment system Tier 1 Design Description in Section 3.2.2.

• Provide heat removal from the spent fuel using the initial inventory of water in the spent fuel pit.

• Provide connections used for temporary emergency makeup to the spent fuel pit.

1he spent fuel pit cooling system is a deferae-in-

! V depth system that removes heat from the spent fuel l stored in the spent fuel pit. Heat removal is accomplished by pumping the water from the spent fuel pit through a heat exchanger, and then returning the water to the pit. The spent fuel pit cooling system is powered from both onsite and offsite ac sources.

The spent fuel pit contains a sufficient volume of water to provide cooling of the spent fuel for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if the spent fuel pit cooling system does not i operate. For cooling beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the spent fuel pit cooling system contains a safety-related connection to

- the spent fuel pit to provide makeup for any water volume loss due to boiling in the spent fuel pit.

i Suction connections to the spent fuel pit are located at an elevation to prevent draining the spent fu-l pit l below a minimum level. The discharge piping to the

_ pent fuel pit prevents siphoning of the pit.

1 3.3.8 1 I

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Tier 1 Design Certification Material SPENT FUEL PIT COOLING SYSTEM Revision: 1 Effective: 12/15/92 . .

Table 3.3.8 Spent Fuel Pit Cooling System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. The spent fuel pit cooling system provides a. Tests shat! be performal to measure the flow a. The flow rate through each heat eschanger heat transfer from the water in the spent rate delivered by each pump to the is 2 [113D] gpm.

fuel pit. associated heat exchanger.

b. Using as-built heat exchanger information b. The calculated heat ren oval capability of and flow rate test data from (l.a.), a heat each heat exchanger is 2 [TBDJ Bruhr.

transfer analysis shall be performed.

i

c. Using as-built elevations and piping c. The calculated pump NPSH is 2 [TBD]

geometry a NPSH analysis for the pumps feet.

shall be performed.

2. The water volume in the spent fuel pit a. A volumetric analysis of the spent fuel pit a. The water volume of the spent fuel pit provides spent fuel cooling for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. shall be performed to verify that the as-built below the spent fuel pit cooling system dimensions are such that the required water pump suction is 2 [T13D] ft'.

solume below the SFS pump suction connection is provided.

b. Inspections shall be perfonrn d to verify that b There are siphon breaks located above the c piping which terminates within the pit below pump suction cor nection on lines the SFS pump suction is provided with terminating below the pump suction siphon breaks above the SFS pump suction cennection.

connection.

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W westinghouse

O Tier 1 Design Certification Material O

SPE'IT FUEL PIT COOLING SYSTEM

, Revision: 1 Effective: 12/15/92 _

Table 3.3.8 Spent Fuel Pit Cooling System inspections. Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests Analyses Acceptance Criteria

3. A makeup connection to the spent fuel A test of the capability to provide rnakeup flow Makeup flow can be delivered to the spent fuel pit is provided. directly to the spent fuel pit shail be conducted pit at 2 [TED} gpm.

using the spent fuel pit cooling system nukeup connection.

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l Tier 1 D: sign Certific: tion M:terial g

L] SERVICE WATER SYSTEM Revision: 1 "1"

Effective: 12/15/92 3.3.0 SERVICE WATER SYSTEM Design Description The service water system (SWS) serves no safety-related function.

'the servlee water system supplies cooling waier to renwve heat from nonsafety related heat exchangers that prmide defense in-depth functions. ' Itis is achieved by pumping service water through the component cooling water heat exchangers.

The service water system provides the defense-in-depth function of removing heat from the component cooling water system for all modes of operation. It removes heat from the spent fuel pool via the spent fuel cooling and component cooling water systems. During plant shutdown, the service water system provides an additional defense-in-depth function of decay heat removal through the nonnal residual heat removal system and the component cooling water system. The

'v service water system is powered from both onsite and offsite se sources. The system includes redundant pumps together with associated piping, valves, and controls.

(D) 3.3.9-1 s W Westinghouse

Tier 1 Design Certification Material SERVICE WATER SYSTEM +

Revision: 1 Effective: 12/15/92 Table 3.3.9 Service Water System Inspections Tests, Analyses and Acceptance Criteria inspections, Tests, Analysis Acceptance Criteria Design Commitment

%e service water system performance shall be Re service water now delivered to one of the

1. The service water system provides validated by a service water system t-st. The component heat exchangers is at least [5000J sufncient How to the component cooling test shan be performed with one of the gpm at minimum suction source !esel as water system o provida the defense-in-redundant pumps eperating, the pump suction derived by analysis.

depth func. .of decay heat removal and source is at least at its minimum eperating level, spent fuel noting.

and with one component coolir,g water heat exchanger and one turbine building closed cooling water heat exchanger in service. His test shall be repeated for the other pump.

De test results shaII be analyzed to assure that the required How is delivered at minimum suction source level. I ne test shall be conducted to verifv on loss of Inspection of the salves confirms that the valves

2. Flowpaths exist to both of the component air or power that a flow path exists to both are in the open position for each of the cooling water heat exchangers with the component cooling water heat exchangers and componmt cooling water heat exchsnger remote turbine building closed cooling water that a now path is isolated for both turbine operated isolation valves and that the remote system heat exchangers iselated on loss of building closed cooling water heat exchangers. operated isolation valves for the turbine air or loss of electrical power.

building closed cooling water heat exchangers are in a closed positien on beth loss of air and foss of power.

3.3.9-2 W Westinghouse

Tier 1 Design Certification Material O

SERVICE WATER SYSTEM , _

Revision: 1 Effective: 12/15/92 _  ;

Table 3.3.9 Service Water System inspections, Tests, Analyses and # ..;eptance Criteria r Design Commitment inspections, Tests, Analysis Acceptance Criteria

3. %e service water system provides A test shall be conducted to verify that the He calculated minimum heat rejection sufficient cooling to the component cooling service water system provides sufficient heat capability of the service water system for water system to perform the defense-in- transfer capab lity at site-limiting environmental defense-in-depth cooling is at least [100 x 10*]

depth functions of normal residual heat conditions. The test can be conducted at any Btu /hr after extrapolating the sneasured value.

removal and spent fuel pool cooling. environmental conditions and the results extrapolated to the limiting condition.

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W Westinghouse

Tier 1 Design Certification Material HYDROGEN CONTROL SYSTEM Revision: 1 Effective: 12/15/92 3.3.10 HYDROGEN CONTROL SYSTEM Desi0n Description ne containment hydrogen control system (VLS) is Defense-in Depth funnions:

a safety-related system consisting of hydrogen control and measurement equipment for the control of

• The hydrogen igniter subsystem controls hydrogen containment hydrogen concentration. He safety-related concentration in excess of the recombiner portions of the hydrogen control system include: capability.

• liydrogen monitoring subsystem consisting of two he power to the hydrogen sensors is provided by independent sets of hydrogen ser. sors distributed the Class IE de and UPS system.

throughout containment. he cables from the containment penetration to the recombiners are Class IE and the terminals for

+ Hydrogen recombiner subsystem consisting of two connecting a temporary power supply to the containment independent recombiners located inside penetration terminal box are Class lE. Power to the containment. He portion of the power supply recombiners is available from both onsite and offsite ac cables for the recombiners located in the sources or from a temporary portable ac power supply.

containment is safety related. He igniters and controls are powered from both onsite and offsite ac sources.

He defense-in-depth portion of the hydrogen ne hydrogen sensors provide an input signal to the control system includes: protection and safety monitoring system.

An actuation signal is manually generated by the

• %e nonsafety-related hydrogen igniter subsystem operator in the main control room at the component consisting of hydrogen igniters distributed level within the plant control system fc,r actuation of the throughout the containment volume. recombiners.

An actuation signal for the igniter system may be -

De hydrogen control system performs the following manually generated within the plant control system in safety related and defense-in-depth functions:. the main control room or within the diverse actuation system.

Safety-related functions:

• He hy dmgen control system provides containment hydrogen concentration measurement.

• Re redundant hydrogen control system hydrogen recombiners comrol contaiament long term hydrogen concentrations.

3.3.10 1 W Westinghouse

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O J Tier 1 Design Certification Material i

HYDROGEN CONTROL SYSTEM Revision: 1 Effective: 12115/92 _

Table 3.3.10 Hydrogen Control System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections Tests, Analyses Acceptance Criteria

1. He hydrogen sensors provide signals to the Simulated hydrogen concentration signals shall Re signals are received by the protection and protection and safety monitoring system. be introduced to demonstrate that the signals are safety monitoring system.

sent to the protection and uferf monitoring system.

2. He VIS hydrogen recombiners provide the Recombiner functional tests shall be conducted Confirm recombiner power can be adjusted to capability to control long term containment to demonstrate the recombiners capability to maintain airflow. A temperature of [1150]

hydrogen concentrations. control containment hydrogen cencentrations. i[TBD]'F within the recombiner heater section is attained with an air flow rate of

[1(X1]i[TBD] scfm.

3. De division assignment for the VIS With the assign-d Class IE power and he hydrogen sensors assigned to the energized hydrogen sensors is such that the loss of protection division energized, tests shall be division receive power.

any single Class IE division does not conducted to confirm power and protection prevent system hydrogen monitoring. division assignments for hydrogen monitoring capability. The tests shall be performed for each assigned power divisioc.

4. The recombiners are energized and heater Recombiner functional tests shall be conducted. a. De hydrogen recombiners are energized power may be varied upon receipt of a upon receipt of an actuation signal.

manual actuation signal.

b. Heater power can be varied from [TBD] to

[00] kw.

. a -2 W westinghouse

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v.s k Tier 1 Design Certificction M:terial v

HYDROGEN CONTROL SYSTEM gj Revision: 1 -

Effective: 12/15/92 Table 3.3.10 Hydrogen Control System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests. Analyses Acceptance Criteria l

5. He VLS hydrogen igniters provide the Igniter functional tests shall be conducted to ne surface temperature of the igniter heating capability to control hydrogen concentration confirm the hydrogen igniters capability to elements is at least [1700]*F.

above the recombiner capability. achieve hydrogen ignition temperature.

6. The igniters are energized upon receipt of Igniter system functional test shall be conducted Each hydrogen igniter is energized upon system an actuation signal from the plant control on the igniters using a manual plant control actuation.

system. system actuation.

7. He igniters are energized upon receipt of Igniter system functional test shall be conducted Each hydrogen igniter is energized upon system an actuation signal from the diverse on the igniters using a manual diverse actuation actuation.

actuation system. signal.

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Tirr 1 D: sign Certific tion M:terirl (m) MAIN AND STARTUP FEEDWATER SYSTEM , , , , ,

(/ Revision: 1  !!!

Effective: 12/15/92 3.4.1 MAIN AND STARTUP FEEDWATER SYSTEM Design Description The main and startup feedwater system (FWS)is a nonsafety-related system which provides the fol6 wing defense-in-depth functions:

• Feedwater is<dation by tripping of the main feedwater pumps.

• Provide startup feedwater for heat removal from the reactor coolant system.

The startup portion of the main and startup feedwater system consists of redundant startup feedw ater pumps, valves, and associated piping. Water is taken from either the deaerator storage tank or the condensate storage tank and is pumped to the steam generators

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using one of two redundant startup feedwater pumps.

Pumps are powered from both onsite and offsite ac sources.

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O Tier 1 Design Certification Material

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5 MAIN AND STARTUP FEEDWATER SYSTEM Revision: 1 Effective: 12/15/92 Table 3.4.1 Main and Startup Feedwater System Inspections, Tests Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests. Analyses Acceptance Criteria I. He startup portion of the feedwater system A functional test shall be performed with each The flow delivered to the steam generator provides flow to the steam generator system of the redundant startup feedwater pumps with system is at least [TBD) gpm at a steam to provide the defense-in-depth function of one of the pump suction sources available. generator secondary pressure of at least [TBD]

heat removal from the reactor coolant psig.

system.

2. De startup feedwater portion of the Pump functional tests shall be conducted he startup feedwater pumps start.

feedwater system stants and operates upon demonstrating pump operation in the presence of receipt of a signal from the plant control an actuation signal. His test shall be performed system. for both pumps.

3. He main feedwater pumps trip upon Pump functional tests shall be conducted ne main feedwater pumps trip.

receipt of a signal from the protection and demonstrating pump trip in the presence of a safety monitoring system. trip signal. This test shall be perforted for both pumps.

4. He condensate storage tank vclume Using as-built dimensions, a volumetric analysis ne tank capacity is at least [TBD) gallons.

provides water for startup feedwater shall be performed to determine the tank volume operation. above the centerline of the startup feedwater ,

pump suction line. >

3.4 1-2 3 Westinghouse

Tier 1 D: sign Certific tion M:terial j l

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[b MAIN STEAM SYSTEM Revision: 1 Effective: 12/15/92 lll l 1

3.4.2 MAIN STEAM SYSTEM Design Description he main steam system (MSS) is a nonsafety-related system which provides the defense inepit; functions of support for main steam isolation and of turbine stop valve closure.

The system provides main steam line isolation using the turbine stop valves, the moisture separator reheater steam supply valves, and the turbine bypass valves for isolation.

De system includes piping from the steam generator system to the condenser and to the turbine stop valves, moisture separator reheater steam supply salves of the main turbine system, and to the turbine bypass valves.

He turbine stop valves receive un actuation signal from both the protection and safety monitoring syr. tem

, and from the diverse actuation system. He main steam

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isolation signal is generated at a system level within the protection and safety monitoring system.

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\, Y Tier 1 Design Certification Material MAIN STEAM SYSTEM -

.g Revision: 1 Effective: 12/15/92 .

Table 3.4.2 Main Steam System inspections, Tcsts, Ann:ysas and Acceptance Criteria Certified Design Commitment inspections. Tests, Analyses Acceptance Criteria

1. Isolation of the steam generator is provided a. Valve functional tests shall be conducted a. He turbine stop valves close, the moisture by the main steam system. demonstrating valve closure in the presence separator reheater steam surrly valves of an actuation signal from the protection close, and the turbine bypass valves close if and safety monitoring system. open.

b. Bypass valve functional tests shall be b. He valves are blocked from opening.

conducted by initiation of an actuation signal from the plant protection and safety rnonitonng system with the turbine bypass valves closed. He operator shall attempt to open the valves.

2. He turbine stop valves close on an actuation Valve functional tests shall be conduct-d he turbine stop valves close.

signal from the diverse actuation system. demonstrating turbine stop valve closure in the l presence of an actuation signal from the diverse actuation system.  ! ,

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Ti r 1 Drsion C:rtific: tion M:ttrial O)

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V DIVERSE ACTUATION SYSTEM Revision: 1 Effective: 12/15/92 3.5.1 DIVERSE ACTUATION SYSTEM Desi0n Description The diverse actuation system (DAS) is a notaafety- actuation system compare these signals to setpoints, related instrumentation and control system that provides When both processors detect a trip condition, a signal to the following defense in-depth functions: cause a reactor trip by deenergir.ing the rod drive motor / generator set. or to actuate an engineered

• Provide automatic actuation signals for a reactor safeguard feature is issued.

trip and a selected set of engineered safeguard Manual actuation of selected plant components is features as a result of certain plant parameters implemented by dedicated switches in the main control exceeding set; mints. room that are wired to these plant components in such a way that the component control path through the

• Provide capability for independent manual protection and safety monitoring system is bypassed.

actuation of a reactor trip and a selected set of The diverse actuation system displays selected engineered safeguard features. sensor outputs in the main control room in a manner that is diverse from the protection and safety monitoring

• Provide independent indication of selected plant system, parameters.

' The diverse actuation system serves no safety-related functions, De hardware and software used to implement the diverse actuation system is diverse from the hardware and software used to implement the protection and safety monitoring system.

The diverse actuation system consists of the signal conditioning, data acquisition, data processors, datalinks and data highways, operator interfaces, display devices, and other equipment necessary for the execution of the functions of the system. The diverse actuation system shares sensor signals with the protection and safety monitoring system and plant control system through isolation devices.

The automatic actuation function of the diverse actuation system is implemented by redundant processors using 2 out-of 2 logie based on sensor inputs.

Sensors in the protection and safety monitoring system and plant control system monitor plant conditions and send signals to the diverse actuation system through isolation devices. The processors in the diverse F 3.5.1 1 i

w W-Westinghouse

O O Tier 1 Design Certification Material 3

I s

DIVERSE ACTUATION SYSTEM @6 Revision: 1 ,

Effective: 12/15/92 Table 3.5.1 Diverse Actuation System Inspections, Tests Analyses and Acceptance Criteria inspections. Tests, Analyses Acceptance Criteria Certified Design Commitment

a. An inspection shall be performed to verify a. The circuit l cards used in the diverse

1. The hardware used to implement the diverse that the electronic circuit boards, used to actuation system are made by a different actuation system is diverse from the implement the diverse actuation system are manufacturer than circuit beards hardware used to implement the protection

{ different from the circuit boards performing performing a similar function in the and safety monitoring system.

similar functions for the protection and protection and safety monitoring system.

safety monitoring system.

b. An inspection shall be performal to verify b. The display devices are made by a different that the display devices uted to implement manufacturer than display desices used in the diverse actuation system are different the protection and safety monitoring from the display devices provided for the system.

protection and safety monitoring system.

c. An inspection shall be performed to verify that any multiplexers used fer the diverse c. Multiplexers are made by a different manufacturer than the multipiexers used in actuation system are different from the multiplexers provided for the protection and the protection and safety monitoring safety monitoring system. system.

An inspection shall be performed to verify that The diverse actuation system software is written

2. The software uwd to implement the diverse the computer code used to implement the diverse using a different computer language.

actuation system is diverse from the software used to implernent the protection actuation system is written in a different and safety monitoring system. computer language from that used fer the protection and safety monitoring system.

l 3.5.1-2 Westinghouse

s A Tier 1 Design Certification MateritJ DIVERSE ACTUATION SYSTEM Revision: 1 h Effective: 12/15/92 '

Table 3.5.1 Diverse Actuation System Inspections. Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests Analyses Acceptance Criteria

3. The diverse actuation system provides Using simulated input signals, system fimetional ne diverse actuation system provides

signals to selected engineered safeguards tests shall be conducted to verify that automatic component actuation signals in respense to features components. and mam al actuation signals are initiated when simulated input signals in acccrdance with system logic has been satisfied. diverse actuation logie.

4. The diverse actuation system prevides Using simulated input signals, system functional The rod drive meter / generator set is manual and automatic means to deenergize tests shaf! be conducted to serify that automatic deenergized in response to simulated input the rod drive motor / generator set. and manual actuation signals deenergize the rod signals in accordance sith diverw actuation l drive motor / generator set when system logic has logic. ,

been satisfied.

5. He diverse actuation system provides An inspection shall be performed to verify that De diverse actuatien system displays the selected plant parameter display in the main the selected plant parameters are displayed in selected plant parameters in the main control control room. the main control room. room.

3.5.1-3 T Westinghouse

Tier 1 DeslDn Certification Material DATA DISPLAY AND PROCESSING SYSTEM Revision: 1 5 Effective: 12/15/92 3.5.2 DATA DISPLAY AND PROCESSING SYSTEM Design Description ne data display and processing system (DDS$

serves no safety-related functions.

The data display and processing system is a nonsafety-relatod instrumentation and control system that provides the display of plant parameters for defense-in-depth functions. The data display and processing system also provides an alternative means of displaying information from the protection and safety monitoring sybtem.

The data display and proccuing system displays this information in the main control room and at the remote shutdown workstation.

He data display and processing system consists of the monitor bus, processors, and displays.

W Westinghouse

's ('N d V Tier 1 Design Certificction Material d

DATA DISPLAY AND PROCESSING SYSTEM Revision: 1 Effective: 12/15/92 Table 3.5.2 Data Display and Processing System inspections, Tests, Analyses and Acceptance Criteria j Certified Design Commitment inspections. Tests, Analyses Acceptance Criteria

1. The data display and processing system An operational test cf the data display and The data display and processing system displays provides equipment to display process processing system shall be performed. in the main control room and at the remote information in the main control rcom and at shutdown workstation are powered and capable the remote shutdown workstation. of disp!aying process data.

a s.2 2 W westinghouse

Ti:r 1 D: sign CertificItion M:ttrirl O INCORE INSTRUMENTATION SYSTEM Revision: 1 Effective: 12/15/92 3.5.3 INCORE INSTRUMENTATION SYSTEM Design Description ne incore instrumentation system (11 S) is a ne thermocouples are positioned within the thimble nonsafety related system except for the following safety- assemblies to measure the local core exit temperature, related functions.

• Provide reactor coolant system pressure boundary integrity

• Provide the protection and safety monitoring system with core exit temperature signals.

He incore instrumentation system includes in core instrument thimble assemblies (which contain fixed core exit thermocouples) associated cabling, electrical con-nectors, and the pressure boundary connection between s the thimble assembly and the guide column surrounding the thimble assembly above the reactor vessel. The core exit thermocouples are installed in the same thimble as the detectors used for the nonsafety-related function of generating a core neutron flux map.

The incore instrument thimble assemblies are inserted into the active core through the upper head and internals of the reactor vessel. The in-core instrument thimble assembly is positioned within the fuel assembly and exits through the reactor vessel head package.

Each incore instrument thimble assembly is composed of instrumentation ar.d wiring including the thermocouple and solid filler material inside a sheath tube. He thimble sheath and pressure boundary litting provide the pressure boundary. The assembly is classi-fled as an instrumentation component.

He pressure boundary fittings are compression type fittings used to provide a seal between the incore instrument thimble and the guide conduit. He fittings permit the exit of the instrumentation cabling and are disassembled every refueling.

l he signals from the core exit thermocouples are used in the protection and safety monitoring system.

3.5.3 1 O)

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Tier 1 Design Certification Matnial iNCORE INSTRUMENTATION SYSTEM Revision: 1 Effective: 12/15/92 Table 3.5.b1 - Incore Instrumentation System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. He incore instrumen' tion system provides Sitaulated signals f hall be introduced at the elec- The thermocouple indicators in the protection core exit thermocouple signals to the trical connectors at the top of the guide tube. and safety monitoring system receive the sig-protection and safety monitoring system. nals. '!

2. He thermocouple is positioned within the Each thimble assembly shall be examined for ne thermocouple position is [TBD] .i [TED]

thimble assembly to measure the local core thermocouple position and tested for electrical from the bottom end of the thimble assembly exit temperature, continuity. and electrical continuity is demonstrated for the thermocouple.

3. He pressure I at.ndary fittings provide ne pressure boundary fittings with temporary The reactor coolant system hydrostatic test is pressure integrity. plugs or caps installed in lieu of thimble assem- completed.

blies snail be inspected during reactor coolant system hydrostatic test.

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Tier 1 Design Certification Material

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PLANT CONTROL SYSTEM . , , , . . . ,

Revision: 1 jf "i[:

Effective: 12/15/92 3.5.4 PLANT CONTROL SYSTEM Design Description The plant control system (P13) is a nonsafety-related system that provides equipment to:

• Validate signals recei/ed from the protection and safety monitoring system to prevent the failure of a sensor or division in the protection and safety monitoring system from propagatin;, into the plant control system.

• Provide ontrol signals for defense-in-depth plant components.

The plant control system consists of sensors, signal conditioning, data acquisition, data processors, datalinks and data highways, and operator interfaces.

E W Westinghouse

,m em (m Tier 1 Design Certification Material l

i PLANT CONTROL SYSTEM [j l Revision: 1 '"

Effective: 12/15/92 ___

i Table 3.5.4 Plant Control System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. He plant control syste n identifies and A functional test shall be performed by The plant control system output signals are

solates failure of process input signals from simulating selected input signal failures. bawd upon the remaining valid input signals.

the protection and safety monitoring system.

? The plant control system provides equipment An operational test of the plant control system The plant control syrem provides control to centrol defense-in-depth plant functions. shall be performed, using selected simulated signals in response to simulated input signals in input signals. accordance with control elgorithms.

W Westinghouse

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l Tilt 1 Daign Certification M:t rill j n

PROTECTION AND SAFETY MONITORING SYSTEM (V) Revision: 1 1

%l Effectivo: 12/15/92 3.5.5 PROTECTION AND SAFETY MONITORING SYSTEM Design Description The protection and safety monitoring system reactor trip or engineered safety features actuation (PMS) provides the following safety-related functions: coincident with a single failure in the PMS, Additionally, the PMS protects against unnecessary

• Tripping the reactor by opening the reactor trip reactor trips or engineered safety features actuations breakers. resulting from single failures in the PMS. loss of power or input signals, or disconnection of portions of

• Actuation of the engineered safety features the system results m a trip or actuation initiating state.

equipment.

Reactor Trip Function

• Safety-related plant parameter monitoring prior to, during, and after an accident or plant transient. 'lhe reactor trip function of the plant monitoring system is implemented by plant sensors, the reactor trip For this design description, the plant monitoring processors, and the reactor trip switchgear. 'ihe reactor system consists o' the sensors, detectors, signal is tripped by opening the circuit breakers in the reactor conditioning, data acquisition, data processors, datalinks trip switchgear, thereby removing electrical power to the and data highwag operator interfaces, displays, and control rod drive mechanisms, causing the control rods V other equipment necessary for the execution of the to drop into the reactor core due to gravity. The reactor functions of the system. The PMS for the APedX) trip breakers are arranged so that tripping any two out implements its functions by software logic installed in of four riivisions results in interruption of nower to the programmable digital devices (data processors). Plant control rod drive mechanisms. Tripping any single data and other signals are exchanged between data division will not interrupt power to the control rod drive processors by means of isolated datalinks and data mechanisms. Once a reactor trip has been initiated, the highways, reactor trip breakers in the reactor trip switchgear latch The sensors and logic for renerating the reactor open, and must be manually rewt before the control trips, enginected safety features actuations, and safety- rmis can be withdrawn.

related plant parameter monitoring are included in the The reactor trip function utilites the four plant monitoring system. PMS components and independent plant monitoring system divisions, using 2-equipment are electrically isolated from nonsafety- out-of-4 logic for automatie trips based on plant sensor related plant instrumentation and electrical equipment. inputa. 'lhe manual reactor trip function uses 1-out-ot-2 Signals from the PMS to other plant instrumentation and logic, control systems, such as the plant control system and the The sensors monitor plant conditions and send data display and processing system, are transa ed signals to the reactor trip processors where these signals through isolation devices. Certain sensor signal, are compared to setpoints. When two or more originating in the PMS are shared with the diverse unbypassed signals monitoring the same plant parameter actuation system through isolation devices. in different divisions exceed the setpoint, and permissive The plant monitoring system is a four division or interlock logic is satisfied, a reactor trip is initiated.

system which automatically or manually initiates a

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Ti:r 1 D: sign Cittific tion Mtt: rial PROTECTION AND SAFETY MONITORING SYSTEM t 4 Revision: 1 Effective: 12/15/92 e

Plant parameters that are monitored to produce a reactor a system level actuation signal is produced in the trip include: engineered safety features actuation processors. This system level signal is transmitted to the associated a Neutron flux protection logie in the same division by the logic bus

• Reactor coolant pump speed data highway. He protection logic then provides

• Reactor coolant flow actuation signals to the component if the component

• Overtemperature AT interlock logic is satisfied. Plant parameters that are

• Overpower AT monitored to produce engineered safety features

• Pressurizer level functions include:

• Pressurizer pressure

• Steam generator level

• Neutron flux

• Reactor trip on safeguards actuation.

• Pressurizer pressure

• Pressurizer level A manual reactor trip and a reactor trip on manual

• Steam generator level safeguards actuation are implemented by directly

• Steam line pressure opening the reactor trip switchgear.

• Cold leg temperature

• Startup feedwater flow Engineered Safety Features Functions

• Containment pressure

• Core makeup tank level.

De engineered safety features functions of the plant monitoring system are implemented by plant The engineered safety features actuation signals sensors, the engineered safety features processors, the include: 1 engineered safety features actuation processors, the protection logic, the logic buses, and manual actuation

• Safeguards actuation devices. The protection logic provides actuating signals

• Passive residual heat removal to operata the plant components. Several engineered

• Core makeup tank injection safety features sensors are shared with the reactor trip

• Reactor coolant pump trip function.

• Containment cooling The engineered safety features functions utilize the

• Containment isolation four independent plant monitoring system divisions.

• Main feedwater line isolation using 2-out-of-4 logic for automatic actuations bued on

• Steam line isolation sensor inputs. An exception is the startup feedwater

• Reactor coolant system signal, which utilizes two divisions and 1-out-of-2 logic. depressurization Manual, systems level actuations are provided for

• Startup feedwater isolation

individual functions.

• Chemical volume control system isolation.

The sensors monitor plant conditions and send

• Turbine trip signals to the engineered safety features processors.

• Steam generator blowdown system isolation where these signals are compared to setpoints. When

• Elock of boron dilution two or more unbypassed signals monitoring the same

• Block steam dump plant parameter in different divisions exceed the a letdown line isolation setpoint, and permissive or interlock logic is satisfied.

• Containment sump pH control

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I l 3.5.5-2 ,

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Tier 1 Design Certification Material PROTECTION AND SAFETY MONITORING SYSTEM E Revision: 1 Effective: 12/15/92

• Normal residual heat removal system isolation.

• Passive residual heat removal heat exchanger outlet temperature Safety-Related Plant Parametsr Monitoring

• Passive residual heat removal flow Function

• Incontainment refueling water storage tank water level The safety-related plant parameter monitoring

• Passive containment cooling flow function is implemented by plant sensors.

• Passive containment cooling storage tank level communications processors or data acquisition

• Containment pressure processors. qualified display processors, and qualit d

• Containment radiation operator displays. Plant sensors may be shared with the

• Containment hydrogen concentration reactor trip and engineered safety features functions. For

• Pressurizer safety valve status phnt sensors shared with either oi these functions, data

• Automatic depressurization system first stage acquisition tak es place at the communications valve status processors, for sensors which are not shared, data

• Automatic depressurization system second stage acquisition is performed by the qualified display data valve status acquisition processors. He plant data is then

• Automatic depressurization system third stage transmitted to the qualified display processors, where it valve status is prepared for display on the qualified operator

• Automatic depressurization system fourth stage displays. valve status.

The safety-tclated plant parameter monitoring function utilizes two of the four independent pimt monitoring system divisions. A minimum of two operator display devices, one per division, are provided at each location. Operator display devices are provided in the nuin control room and at the remote shutdown workstation.

The sensors monitor plant conditions and send _

signals to either the communications processors, or the qualified display data acquisition processors. His data is transmitted to the qualified display processors, where it is collected, organized, and prepared for display. The final data is displayed on the qualified operator displays.

The phmt parameters that are collected and displayed by the safety-related plant parameter monitoring function include:

• Reactor coolant system pressure

• Reactor coolant system temperature

• Pressurizer level

• Neutron flux

• Containment water level

• Core exit temperature 3.5.5-3 W-Westinghouse

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v v v Tier 1 Design Certification Material PROTECTION AND SAFETY MONITORING SYSTGa Revision: 1 Effective: 12/15/92 r

I Table 3.5.5 Protection and Safety Monitoring System inspections, Tests, Analyses and Acceptance Criteria i

Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria l

1. De protection and safety monitoring system a. System functional tests shall be conducted to a. Reacter trip breakers open when trip logic performs the safety-related reactor trip, verify that reactor trip breakers open when is satisfied from the following plant engir,eered safety features actuation, and system logie has been satisfied. parameters:

plant parameter monitoring functions.

• Neutron flux

• Reactor coolant pump speed

• Reactor coolant flow

= Overtemperature AT

= Overpower AT

• Pressurizer level

= Pressurizer pressure

• Steam generator level.

W Westinghouse

Tier 1 Design Certification Material

[jf.

PROTECTION AND SAFETY MONITORING SYSTEM F Revision: 1 ,

Effective: 12/15/92 Table 3.5,5 Protection and Safety Monitoring System inspections, Tests, Analyses and Acceptance Criteria inspections, Tests, Analyses Acceptance Criteria Certified Design Commitment

b. System functional tests shall be conducted to b. Component actuation signals are generated

1. (continued) when engineered safety features actuation verify that engineered safety features actuation signals are initiated when system logic is satisfied from the following plant logic has been satisfied. parameters:

= Neutron flux

• Pressurizer pressure

• Pressurizer level

• Steam generator level

• Steam line pressure

• Cold leg temperature

• Startup feedwater flow

• Containment pressure

• Core makeup tank level.

1 3.5.5-5 W Westinghouse

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Q Tier 1 Design Certification Material PROTECTION AND SAFETY MONITORING SYSTEM Revision: 1 Effective: 12/15/92 .

Table 3.5.5 Protection and Safety Monitoring System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. (continued) c. An inspection shall be performed to verify c. The protection and safety monitoring system that the designated plant parameters are displays the following plant parameters in displayed. the main control room and at the remote shutdown workstation:

• Reactor coolant system pressure

• Reactor coolant system temperature

• Pressurizer level

• Neutron flux

• Containment water level e Core exit temperature

• Passive residual heat removal system heat exchanger outlet temperature

• Passive residual heat removal flow

• Incontainment refueling water storage tank water level

• Passive containment cooling flow

• Passive containment cooling storage tank level

• Containment pressure

• Containment radiation

• Containment hydrogen concentration W-Westinghouse

O O Tier 1 Design Certification Material O

PROTECTION AND SAFETY MONITORING SYSTEM Revision: 1 l!l Effective: 12/15/92 ,

F Table 3.5.5 Protection and Safety Monitoring System Inspections, Tests, Analyses and Acceptance Criteria 1

l Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. (continued)

• Pressurizer safety valve status

• ADS system first stage valve status

• ADS second stage valve status

• ADS third stage valve status

• ADS fourth stage valve status.

d. System functional tests shall be conducted to d. Operational permissives and interlocks are serify that operational permissives and generated and removed when reactor trip interlocks are generated and removed when and engineered safety features actuation system logic has been satisfied. logic is satisfied from the following plant parameters:

• Neutron Hus

• Pressurizer pressure.

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y/ s Tier 1 Design Certification Material v

PROTECTION AND SAFETY MONITORING SYSTEM Revision: 1 Effective: 12/15/92 .

Table 3.5.5 Protection and Safety Monitoring System Inspections, Tests, Analyses and Acceptance Criteria m

Certified Design Commitment inspections. Tests, Analyses Acceptance Criteria

2. The protection and safety monitoring a. Tests shall be conducted to measure the a. De time to satisfy trip logie, the trip signal system design provides timely initiation of response times to initiate reactor trip when to reach the reactor trip breakers, and the safety-related reactor trip and engineered trip setroints have been exceeded. reactor trip breakers to open is less than or safety features actuations. equal to the time response requirement Time response is defined as the maximum listed for the following channels:

allowable time for the reactor trip breakers to open following a step change by a

• Power range neutron flux s [TBD] see simulated sensor from 5% below the

• Reactor coolant pump speed setpoint to 5% above the seipoint with each 6 [TBD] see externally adjustable time deia, set to OFF.

• Reactor coolant flow s [TBD] sec

• Overtemperature AT s [TBD] sec

= Overpower AT s [TBD] sec

• Pressurizer level s [TBD] sec

= Pressurizer pressure s [TBD] sec

• Steam generator narrow range level s [TBD) sec.

W Westinghouse l

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Nj V Tier 1 Design Certification Material PROTECTION AND SAFETY MONITORING SYSTEM _

Revision: 1 Effective: 12/15/92 _

Table 3.5.5 Protection and Safety Monitoring System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

2. (continued) b. Tests shall be conducted to measure the b. The time to satisfy engineered safety response times to initiate engineered safety features actuation logie and the component features actuation signals when trip setpoints actuation signal :o be produced is less than have been exceeded. or equal to the time response requirement listed for the following channels:

Time response is defined as the maximum allowable time for component actuation

• Source range neutron flux (rate) signals to be produced following a step s [TBD] sec change by a simulated sensor from 5%

• Pressurizer pressure s [TBD] see below the setpoint to 5% above the serpoint

• Pressurizer level s [TBD] see with each externally adjustable time delay

• Steam generator narrow range level set to OFF. Time respcmse shall not s [TBD] see include the engineered safety features

• Steam generator wide range leve!

components. s [TBD] sec

• Steam line pressure s [TBD] sec

• Cold leg temperature s [TBD] sec

• Startup feedwater flow s [TBD] sec

• Containnent pressure s [TBD] sec g

• Core makeup tank level s [TBD] sec.

W Westinghouse

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Tier 1 Design dNtification Materici v)

PROTECTION AND SAFETY MONITORING SYSTEM Revision: 1 Effective: 12/15i92 _ .I Table 3.5.5 Protection and Safety Monitoring System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria 3.a. He protection and safety monitoring a. He manual reactor trip switches shall be a. He reactor trip breakers open when the system provides a manual reactor trip tested. manual reactor trip switches are operated.

capability.

b. The protection and safety monitoring b. He manual safeguards actuation switches b. He reactor trip breakers open when the system initiates a reactor trip coincident shall be tested. manual safeguards actuation switches are with manual safeguards actuation. operated.

c. He protection and safety monitoring c. He fo!!owing manual engineered safety c. Component actuation signals are generated system provides manual engineered safety features actuation switches shall be tested: in accordance with engineered safety features actuation capability. features actuation logic when manual

• Manual safeguards actuation engineered safety features actuation

• Manual passive residual heat removal switches are operated.

actuation

• Manual steam line isolation

• Manual steam /feedwater isolation

• Manual feedwater isolation

• Manual containment cooling actuation

• Manual containtnent isolation actuation

• Manual depressurization system actuation.

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Tier 1 Design Certification Material l

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PROTECTION AND SAFETY MONITORING SYSTEM l Revision: 1 j l

Effective: 12/15/92 Table 3.5.5 Protection and Safety Monitoring System l Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests. Analyses Acceptance Criteria 1

l

4. He four reduralant divisions of protection One protection and safety monitoring system %e acceptance criteria are the same as the  ;

1 and safety monitoring system equipment are division shall be selected and deenergized. He acceptance criteria for the Inspections. Tests. l independent from each other except for tests of the Inspections, Tests, Analysis, and Analyses and Acceptance Criterin, l isolated data communications required for Acceptance Criteria, Table 3.5.5-1, entries 1(a), Table 3.5.5-1, entries f(a),1(b),1(c), and 1(d) voting logic. The four redundant divisions 1(b),1(c), and 1(d) shall be repeated. except for the division that is deenergized.

of protection and safety monitoring system equipment are powered from independent power sources.

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Ti:r 1 D: sign C:rtific tion M tiri:1

/m RADIATION MONITORING SYSTEM (U} Revision: 1

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Effective: 12/15/92

. i 3.b.6 RADIATION MONITORING SYSTEM Design Description The radiation monitoring system (RMS) is a

• Measure and record the concentration of radioactive nonsafety-related system except for those portions w hich materials in the liquid discharge to the environment perform the following safety-related functions: and provide alarm and indication in the main control room.

• Measure the radioactivity levels in the containment atmosphere and provide signals to the protection and safety monitoring system

• Measure the concentration of radioactivity in the main control room normal supply air and provide a ignal to the protection and safety monitoring system.

,e~ Power for the safety-related portions of radiation

(' monitoring system is provided from the Class IE de and UPS system. The division assignment is such that the loss of any single Class IE power division will not prevent the system from accomplishing its safety-related functions.

He radiation monitoring system performs the following defense-in-depth functions:

• Measure the concentration of radioactivity in the main control room normal supply air and provide a signal to the plant control system

• Measure, display and record the concentration of radioactive materials in the plant vent efuuent to the atmosphere and provides alamts and indication in the main control room

• Measure, display and record the amount of radioactivity in the condenser air removal discharge path to the atmosphere and provide alarnw and indication in the nuin control room O

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Tier 1 Design Certification Material RADIATION MONITORING SYSTEM Revision: 1 .,

Effective: 12/15/92 Teble 3.5.6 Radiation Monitoring System Inspections, Tests, Analyses and Acceptance Criteria inspections, Tests, Analyses Acceptance Criteria Certified Design Commitment Using simulated signals, functional tests shall be %e signals are received by the protection and

1. He containment high range monitor provides performed to demonstrate that the radiation signals safety monitoring system.

radiation level signals to the protection and safety monitoring system. are sent to the protectica and safety monitoring system.

1

\

Using a simulated signal, a functional test sha!! be ne signal is received by the pmtection and

2. The control room supply air duct monitor performed to demonstrate that the radiation signal safety monitoring system.

.nrovides a radiation level signal to the protection and safety monitoring system. is sent to the protection and safety monitoring system.

Using a simulated signal, a functional test shall be The signal is received by the plant control

3. He control room supply air duct monitor performed to demonstrate that the radiation signal system.

provides a rediation level signal to the plant centrol system. is sent to the plant control system.

Using simulated signals, functional tests shall be ne sigul is received in the main control

4. He plant vent monitor records and displays performed to demonstrate that the radiation signals room.

the plant vent effluent radiation levels.

are provided to the main control room.

Using simulated signals, functional tes:s shall be De signal is received in the main control

5. He condenser air removal discharge monitor performed to demonstrate that the radiation signals room.

records and displays condenser air removal effluent radiation level. are provided to the main control room.

3.5.6-2 W-Westinghouse

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Tier 1 Design Certification Material L)

RADIATION MONITORING SYSTEM Revision: 1 Effective: 12/15/92 _ e Table 3.5.6 Radiation Monitoring System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

6. He liquid radwaste discharge monitor displays Using a simulated signal a functional test shall be ne signal is received in the main control the radwaste discharge radiation levels, performed to demonstrate that the radiation signal room.

is provided to the main control room.

7. He division assignment for the safety-related With only the assigned Class IE power and ne RMS channels assigned to the energized RMS channels is such that the loss of any protection division energized, tests shall be division have power available.

single Class IE division will not prevent the performed to confirm power and protection system from measuring the radioactivity levels division assignments by operating the channels.

in the containment atmosphere and the main His test shall be perfonned for each assigned control room and provide these signals to the power division, protection and safety monitoring system.

W-Westinohousea

Ti;r 1 D; sign Certific; tion M;t: rial i

7.

! ) MAIN AC POWER SYSTEM U' Revision: 1 ,

Effective: 12/15/92

. e 3.6.1 MAIN AC POWER SYSTEM Design Description The main ac power system (ECS) serves no safety-related function with the exception of intenuption of the power to the reactor coolant pump motors.

The safety-related function of interrupting the power to reactor coolant pump motors is accomplished by two Class IE breakers in series for :ach reactor coolant pump. The control power to these Class IE breakers is provided by the Class IE de and UPS System.

Yhe nuin ac power system perfonns the defense-in-depth function of distributing non-Class IE ac power from offsite and onsite sources to selected components in order to accomplish defense-in-depth functions.

The defense-in-depth nonsafety-related loads are divided into two functionally redundant load groups and (O

C/

are powered from the sources that include startup transfonner system, onsite standby power system and the reserve auxiliary transformer. The main ac power system provides medium and low voltage power to the associated defense-in-depth loads.

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V) l (s Tier 1 Design Certification Material

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MAIN AC POWER SYSTEM 7 Revision: 1 Effective: 12/15!92 .

Table 3.6.1 Main AC Power System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. The reactor coolant pump breakers open Ilreaker functional tests shall be conducted Reactor coolant pump breakers open upon receipt upon receipt of an actuation signal from the demonstrating breaker operation upon receipt of of an actuation signal from the protection and protection and safety monitoring system. an actuation signal. safety monitoring system.

2. The divisional assignment for the reactor With only the assigned Class 1E power and ne two breakers associated with each reactor coolant pump breaker controls is such that protection division energized, tests shall be coolant pump are controlled by different Class the loss of single Class iL division will not performed to confirm power and protection 1E divisions.

prevent interruption of power to each division arrangement by operating the reactor reactor coolant pump. coolant pump breakers. His test shall be performed for all assigned power divisions.

3. Power is provided to the functionally Inspection and test shall be performed to verify Each of the component redundant loads is redundant defense-in-depth load groups that each of the component redundant loads is connected to a separate ac bus, from separate ac buses, assigned to its associated bus.

4. Each of the defense-in-depth load groups is A test shall be performed to verify that the ne desired ac voltage and frequency are powered from the startup transformer assigned bus for each load group may be powered available for each load group bus power source system, the onsite standby power system, from the startup transformer system, onsite option.

and the reserve auxiliary transformer. standby power system, and the reserve auxiliary transformer.

'6 W- WestinEhouse

Ti:r 1 DIsign Ctrtifiertion Mit: rial n

NON CLASS 1E DC AND UPS SYSTEM (V) Revision: 1 5I Effective: 12/15/92 3.6.2 NON-Cl. ASS 1E DC AND UPS SYSTEM Design Description The non-class IE de and UPS system (EDS) serves no safety-related function. The EDS systerr.

provides electrical power for control and monitoring of the defense-in<lepth functions.

The EDS system is comprised of two subsystems representing two power load groups. Each subsystem consists of two sets of battery chargers, stationary battery banks, inverters, regulating transformers, and the associated distribution equipment.

During normal operation, the battery chargers supply the continuous de load demands while maintaining the associated battery banks fully charged.

The uninterruptible ac power is normally provided by the inverter. The associated regulating transformer provides the backup ac power should the inverter be unavailable.

) Each hattery bank supports its associated defense-(O in-depth loads for a duration of at least two hours during loss of all ac power sources.

3.6.2-1 (D) v W-Westinghouse

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k s Tier 1 Design Certification Material NON-CLASS 1E DC AND UPS SYSTEM ig Revision: 1 Effective: 12/15/92 _

Table 3.6.2 Non-Class 1E DC and UPS System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections. Tests, Analyses Acceptance Criteria ,

I. Each battery bank in a non-class IE de An eight hour constant current capacity test at At the end of the test, battery terminal voltage subsystem meets its design load [478 + 0.5%] amperes shall be performed on is greater than or equal to [105] volts.

requirements without the support of each battery bank to demonstrate that the battery battery chargers for at least two hours. bank is capable of meeting its design load requirement.

2. Each non-Class IE battery charger meets A load test shall be performed to demonstrate Each battery charger capacity is at least [1000]

the continuous load demand w hile that each battery charger is capable of meeting de amperes.

maintaining the associated battery bank the continuous load demand while maintaining fully charged. the associated battery bank fully charged.

3. Each non-Class IE inverter provides A load test shall be performed to demonstrate Each inverter capacity is at least [75} kva.

uninterruptible ac power to the selected that each inverter is capable of supplying non-Class lE instrument and control uninterruptible ac power to the non-Class 1E equipment required to support defense-in- instrumentation and control load required to depth functions. support defense-in-depth functions.

4. Each non-Class IE regulating transformer A load test should be performed to demonstrate Each regulating transformer capacity is at least provides ac power to the non-Class IE that each regulating transformer is capable of [75] kva.

instrument and control loads normally supplying the inverter loads.

supplied by its associated inverter.

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Tirr 1 D: sign C:rtific: tion M:t: rial

[ h PLANT LIGHTING SYSTEM

'v' Revision: 1 E Effective: 12/15/92 3.6.3 PLANT LIGHTING SYSTEM Design Description The plant lighting system (ELS) performs no safety-related function. 'ne emergency lighting in the main control room and the remote shutdown workstation area performs defense-in-depth function by providing illumination for emergency operations upon loss of nonnal lighting.

Main control room and remote shutdown workstation area emergency lighting power is supplied from the Class IE de and UPS system.

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Tier 1 Design Certification Material PLANT LIGHTING SYSTEM w

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Revision: 1 Effective: 12/15/92 _

e Table 3.6.3-1, Plant Lighting System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. The emergency lighting in the main control room A functional test shall be performed by Illumination level of [10] footcandles measured and the remote shutdown workstation area deenergizing the normal lighting in the main three feet from the floor is available to two provides illumination to two main control room control room and the remote shutdown main control room workstations and the workstations and the remote shutdown workstation area. 'The illumination level for remote shutdown workstation.

workstation. emergency lighting shall be measured.

W Westinghouse

Titt 1 Dzsign Ccrtific: tion M::t: rial n

(d T CLASS 1E DC AND UPS SYSTEM Revision: 1 Effective: 12/15/92 3.6.4 CLASS 1E DC AND UPS SYSTEM Design Description ne Class IE de and UPS System ODS) provides The safety-related functions of the battery banks, power for the safety-related equipment, battery chargers, UPS inverters, and the regulating There are four independent Class IE 125 vde transformers are:

divisions, A, B. C, and D. Divisions A and D are each comprised of one battery bank, one switchboard, and

• He battery banks provide de power to the safe one battery charger. Divisions B and C are each shutdown loads for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Divisions A and D) comprised of two battery banks, two switchboards, and and for 24 and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (Divisions B and C) two battery chargers. respectively, as required, without the support of One battery bank in each division, designated as the battery chargers during a loss of all ac power 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> battery bank, provides power to the loads sources, required for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following an event of loss of all ac power sources. The second battery bank in

• De battery chargers provide electrical isolation divisions B and C, designated as 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> battery bank, between the main ac power system and Class IE is used for those loads requiring power for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 125 vde circuits.

following the same event, h} De Class IE UPS provides power to four

• The UPS inverters provide ac power to the V independent divisions of Class I E instrument and control instrument and control devices, and the emergency buses and emergency lighting. Divisions A and D cach lighting.

consist of one Class IE inverter associated with an instrument and control distribution panel and a backup

• The regulating transformers provide electrical regulating transformer. The inverter is powered from isolation between the main ac power system and the respective 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> battery bank switchboard. Class IE UPS circuits.

Divisions B and C, each consists of two inverters, two instrument and control distribution panels, and a backup In addition to the above safety-related functions, the regulating transformer with a distribution panel. One battery chargers and the regulating transformers perform inverter in Division B and C is powered by the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the following defense-in-depth functions:

battery bank switchboard and the other by the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> battery bank switchboard.

• The battery chargers supply the continuous load During normal operation, the Class 1E inverters demand while maintaining the associated batteries in receive power from the associated de bus. If an inverter charged condition.

is not available, the Class IE regulating transformer provides backup power to Class IE UPS loads from the

• The regulating transformer in each division main ac power system. provides backup power to the Class IE UPS loads ne Class IE de and UPS system is designed so if one of the inverters in that division is not that no single failure of any component results in loss of available.

more than one Division.

O W Westinghouse 3.G.4-1

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N./ %s J Tier 1 Design Certification Material CLASS 1E DC AND UPS SYSTEM Revision: 1 Effective: 12/15!92 _

Table 3.6.4 Class 1E DC and UPS System inspections, Tests, Analyses and Acceptance Criteria Certified De ign Commitment inspections, Tests, Analyses Acceptance Criteria

1. Each battery bank in a Class IE Division An 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> constant current capacity test at [500] At the end of the test, battery terminal voltage is meets its design load requirements without amperes shall be performed on each 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and greater than or equal to [105] volts.

the support of battery chargers for 24 or 72 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> battery bank to demonstrate that the hours, as required. battery bank is capable of meeting its design load requirement.

2. Each Class iE battery charger meets the lead test shall be performed to demonstrate that Each battery charger capacity is at least [240]

continuous load demand while maintaining each battery charger is capable of meeting the amperes.

the associated battery bank charged. 03rainuous load demand while maintaining the associated battery bank charged.

3. Each Class IE inverter provides power to the Imad test shall be performed to demonstrate that Each inverter capacity is at least [10] kva.

Class iE instrument and control loads of its each inverter is capable of meeting the Class 1E safety Division. instrument and control loads of its safety division.

4. Each Class IE regulating transformer lead test shall be performed to demonstrate that Each regulating transformer capacity is at least provides power to the Class lE instrument each regulating transformer is capable of meeting [20] kva.

and control loads of any inverter in its safety the Class 1E instrument and control loads of the Division. inverter in its safety Division.

W Westinghouse

Tier 1 Design Certification Material ONSITE STANDBY POWER SYSTEM . . . . . .

Revision: 1 55' Effective: 12/15/92 3.6.5 ONSITE STANDBY POWER SYSTEM Design Description The Onsite Standby Power System (ZOS) serves Each diesel generator supplies power at the no safety-related function. ZOS supplies ac power to nominal voltage and frequency ratings of the plant the following systems loads that are automatically medium voltage bus.

connected to the diesel buses in a predetermined time sequence:

a. the Class IE Direct Current and UPS System,

b. the selected electrical components of the plant defense-in-depth non safety-related systems.

Thee systems loads are divided into two functionally redundant load groups each supplied by a separate diesel generator.

ZOS includes two onsite standby diesel generator units that supply power to the associated main ac power system distribution buses. Each diesel generator unit is complete with its own support subsystems that include:

• Diesel Engine Starting Subsystem ~

• Combustion Air Intake and Engine Exhaust Subsystem

• Engine Cooling Subsystem

• Engine Lubricating Oil Subsystem

• Engine Speed Control Subsystem

• Static exciter, generator protection, monitoring instruments and control subsystems.

Each diesel-generator unit has capacity to power selected system components.

Each diesel-generator unit is capable of automatically starting and connecting to its associated bus in the event of a loss of voltage on these buses.

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\ , i e (U V Tier 1 Design Certification Material U

ONSITE STANDBY POWER SYSTEM Revision: 1 Effective: 12/15/92 .

Table 3.6.5 Onsite Standby Power System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. In the event of loss of voltage on the Each diesel generator shall automatically start On receipt of loss of D/G backed bus voltage diesel generator backed buses, each diesel and connects to its associated bus in the event of signal, each of the diesel generator starts and generator automatically starts and loss of voltage on that bus. powers the associated bus connects to its associated bus.

He time from the instant of loss of bus voltage

• at the rated [60 0.33 %] hz frequency signal to the instant rated voltage reappears on and [4160 i 0.5 %) volts voltage the bus shall be measured.

• w ithin [20] seconds.

2. He selected systems loads are A test shall be performed to verify that each he test results verify that the selected systems automatically connected to the diesel automatic load sequencer connects the selected load breakers close in a predetermined time backed buses per predetermined loading systems loads to the associated bus in a sequence.

sequence. predetermined time sequence.

This test can be performed by verifymg the individualload breaker operating in TEST position.

3. Each diesel-generator operates at the load He diesel generator shall be operated at ne test results serify that ech diesel generator capacity required to supply the selected minimum [2736] kw capacity for a time required can support the plant defense-in-depth loads defense-in-depth systems loads. to reach engine temperature equilibrium plus continuously without exceeding the design one hour. temperature limits.

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'p 'N A O (V Tier 1 Design Certification Material C),

ONSITE STANDBY POWER SYSTEM - -

Revision: 1 Effective: 12/15/92 ,

Table 3.G.5 Onsite Standby Power System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment Inspections, Tests, Analyses Acceptance Criteria

4. The diesel generator governor and A test shall be performed either in the vendor The test results and the associated analysis excitation / voltage regulator can provide shop or in the field to measure the voltage and confirm that the bus voltage is maintained voltage regulation and frequency control for frequency readings when: above [4160 - 20%, and + 10%] volts and the the defense-in-depth loads during automatic frequency is maintained above [60 - 5%,

load sequencing. a. the diesel generator is operating at no load and + 2%] lh during lead sequencing.

and is subject to a stepload of [TBD] kw,

,. b. the diesel generator is operating at [TBD]

P percent capacity and is subjected to a stepload of [TBD] st [TBD] seconds,

c. Repeat the steps as necessary to simulate the largest incremental loading at time interval

[TBD] seconds.

An analysis of the shop / field tests shall be performed to ensure that the test conditions envelop the automatic load sequence profile.

3.G.5-3 W Westinghuse

Ti:t 1 Dstign C:rtification M:ttriti

,.~

I NUCLEAR ISLAND NONRADIOACTIVE VENTILATION SYSTEM

\ Hevision: 1 Effectivo: 12/15/02 _

3.7.1 NUCLEAR ISLANI) NONRADIDACTIVE VENTILATION SYS'UM Desi0n Description

'Ihe nuclear istand nonradioactive ventilation systr a provided by a supplemental air filtration unit which (Vils) is a nonsafety-related system except for provides llEPA filtration, components that provide isolation of the main control The nuclear island nonradioactive ventilation system --

room envelope. is powered from t oth onsite and offsite ac powet

'lhe safety-related function of main cont ul room sources.

envelefe isola ..on is provided by redundant main control room iholatic,e dampers and anociated ductwork that .

maintain the integrity of the main control room em elope. The power supply to the safety-relued

, isolation dampers is provided by Class IE de and UPS system. The safetyclated function of providing cooling to the main cor. trol room and Class IE electrical rooms is provided by the main control room emergency habitability system. Refer to the main control room emergency habitability system Tier 1 Design Des-ription

(

, in Section 3.2.6.

The defense in. depth functions of the nuclear island nonradioactive ventilation system are to:

b*

+ Provide ventilation and cooling to the main control room envelope, Class IE instrumentation and control rooms Class IE de equipment rooms, and -

Class lE battery roome.

• 1.imit the airborne radioactivity level within the main control room envelope.

The de fense-in<terth function of providing ventii. tion and cooling for the main control room envelope, Class IE instrumantation and control rooms.

Class IE de equipment rooms, and Class IE battery ams is provided by redundant air handling units, return / exhaust fans, dampers, and the associated ductwork distribution system.

The defense-in-depth function of limiting airborne radioactivity level within the main control room is

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'N 3.7.1-1 W Westinghouse Q

%) v Tier 1 Design Certification Material NUCLEAR ISLAND NONRADIOACTIVE VENTILATION SYSTEM 9 Revision: 1 Effective: 12/15/92 Table 3.7.1 Nuclear Island Nonradioacti"e Ventilation System inspections. Tests Analyses and Act.eptance Criteria Certified Design Commitment inspections. Tests, Analyses Acceptance Criteria L

1. He rmin control room envelope isolation Damper functional test shall be conducted to ne ma:n control room envelepe isolation dampers close upon receir* of an actuation demonstrate damper operation upon receipt of damper < close upon receipt of an actuation signal from the protection and safety an actuation signal. signal.

monitoring system.

2. The division assignment for the main control With only the assigned Class 1E power and he dampers assigned to the energized divisien room isolation dampers and controls is such protxtion division energized, tests shall be close.

that the loss of rny single Class 1E division performed to confirm power and protection will not prevent the system from division assignment by operating the dampers accomplishing its safety-related function. His test shall be performed for all assigned power divisions.

3. He main control room HVAC subsystem Supplemental air filtration unit HEPA 61ters Supplemental air filtration units HEPA filters supplemental filtration unit limits the airborne shall be functionally tested to verify removal provide a [99%] removal efficiency for radioactivity by use of HEPA filtus. efficiency. particulate.

4. He main control room HVAC < absystem air Air handling unit high efficiency filte:s shall be Main control room air handling unit high handling units limit the airborrn particulate functionally tested to verify particulate removal efficiency filters provide an [S04] removal by use of high efficiency filters. efficiency. efficiency for airborne particulate.

5. He main control room HVAC subsystem System functional tests shall be conducted to The main control reom envehr ambient supplementi air filmtion unit provides demonstrate the system capability to achieve a pressure is at icast [I/8] inch water gauge greater makeup air *-J mamtains the main control positive pressure in the main control room than the surrounding areas.

room envelope at positive pressure with envelope when operating with the supplemeotal respect to the surro mding areas. air filtration unit providing makeup air.

W Westinghouse

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G J Tier 1 Design Confication Material G

NUCLEAR ISLAND NONRADIOACTIVE VENTILATION SYSTEM iiE =

Revision: 1 Effective: 12/15/92 Table 3.7.1 Nuclear Island Nonradioactive Ventilation System inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests. Analyses Acceptance Criteria

6. He main control room HVAC subsystem Using simulat-d signals, a functional test shall ne supplemental air filtration unit starts and realigns to the supplemental air filtration be conducted to demonstrate the system realigns dampers realign.

mode upon receipt of a plant con'rol system to the supplemental air filtration mode.

actuation signal cf high radiation in the supply air.

'7. He VB5 provides ventilation and cooling to System functional tests sha!! be conducted to he mair centrol room envelepe and Class 1E main control room envelope, Class 1E demonstrate proper system eperation to maintain instrume atation and control rooms. Class 1E de instrumentation and control rooms, Class 1E room temperatures within maximum limits. equipm .nt rooms, and Class IE battery froms de equipment rooms, and Class IE battery temperatures are maintained less than or equal rooms. to [75]* F.

3'7'

W-Westinghouse

i Ti:r 1 D: sign C:rtific: tion M;t: rid l CENTRAL CHILLED WATER SYSTEM Revision: 1 .

Effective: 12/15/92 e

3.7.2 CENTRAL CHILLED WATER SYSTEM Design Description 1he central chilled water system (VWS) serves no safety-related functions other than containment inla. tion.

The containment isolation safety-related function is covered in the containment system Tier i Design Description in Section 3.2.2.

'Ihe low capacity subsystem of the central chilled water system performs the defense.in-depth function of supplying chilled water to support the following cooling functions:

• Nuclear istsnd non-radioactive ventilation system cooling for the main control room envelope

• Vils cooling for the Class IE instrumentation and control reonw, Class IE de equipment rooms, and Class IE battery rooms v

• Compartment unit coolers for the normal residual heat removal system pump

• Compartment unit coolers for the chemical and volume centrol system makeup pump.

'the low capacity subsystem consists of redundant pumps, air-cooled chillers, and associated piping and valves. The low capacity subsystem is powered from both onsite and offsite ac power sources, r 3.7.2 1

( W-Westinghouse s

Tier 1 Design Certification f.1sterial CENTRAL CHILLED WATER SYSTEr4 IM Revision: 1 Effective: 12/15/32 Table 3.7.2 Central Chilled Water System inspections, Tests, Analyses and Acceptance Criteria f Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria He low capacity subsystem performs the A test of the low capacity subsystem shall be a. He chilled water flow rate is not less than 1.

performed to measure the chilled water flow rate [560] gpm for each of the redundant defense-indepth function of supplying chilled and supply tempe ature. subn stems.

water to support area cooling of the main ,

control room. Class IE instrumentation and l

b. De chilled water supply temperature is not control rooms, Class IE de equipment rooms, Class !E battery rooms. and the greater than [42]' F.

normal residual heat removal and ch mical and volume control makeup pump compartments. _._

A functional test shall be conducted to Each unit cooler starts automatically in response

2. The normal residual heat removal and chemical and volume control pump room demonstrate that the normal residual hest to pump eperation or high roem temperature removal and chemical and volume control pump signals and maintains the normal residual heat unit coolers maintain acceptable room unit coolers automatical?y start whenever removal and chemical and volume control compartment temperatures during pump the associated pump is started or upon reaching compartment temperature at or below [130}' F operation.

a high room temperature of {110}' F and during pump operation.

provide cooling to the recms. l l

3 Westinghouse

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Ti:r 1 D: sign Csrtification M:t:ri:1

{s ANNEX / AUX BUILDINGS NONRAD10 ACTIVE VENTil.ATION GYSTEM Revision: 1 l

Effectivn: 12/15/92 l

l 3.7.3 ANNEX / AUX BUILDINGS NONRADIOACTIVE VENTILATION SYSTEM l

Design Description  !

1he annexiauxiliary buildings nonradioactive ventilation system (VXS) serves no safety related function. The equipment room flVAC subsystem provides ventilation of the electrical switchgear rooms that contain the diesel but switchgear. Ventilation and emling of these diesel bus switchgear is required for the onsite standby power system to perform its defense-in-depth functions.

The equipment rooir HVAC i,ubsystem consists of two supply air handling units and two return / exhaust fans w hich maintain the sitchgear rooms below the design maximum temperature w hen the onsite standby power systern is operating during a loss of offsite pow er.

Isolation dampers are provided in the recirculation

, [ oucts between the return air fans and the supply air

\ handling units. These dampers close on a loss of power to align the sys,tein to a once through ventilation mode.

Isolation dampers are also provided in selected duct branches to other building areas. These dampers close on a loss of pow er and a signr.1 that a fan has no flow.

Closure of these dampers directs system flow to the switchgear rooms.

1 A 3.7.3-1 W Westinghouse V) -

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O Tier 1 Design Certification Material O

ANNEX / AUX BUILDINGS NONRADIOACTIVE VENTILATION SYSTEM M F

Revision: 1 Effective: 12/15/92 Table 3.7.3 Annex! Auxiliary Buildings Nonradioactive Ventilation System Inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment Inspections, Tests, Analyses Acceptance Criteria

1. When the onsite standby power system s) A test shall be performed to simulate the a) Outside air intake and exhaust dampers are operates during a loss of of fsite power, the loss of offsite power. open.

electrical switchgear rooms are maintained at or below the design maximum Measure air flow to and from each diese! Air handling unit fans and return fans are temperature, generator bus suitchgear room. eperating. j l

Recirculation duct isolation dampers are closed.

A minimum flow of [12AKM) scfm] is supplied and exhausted from each switchgear rcem.

b) A test shall be performed to simulate the b) Outside air intake and exhaust dampers are loss of offsite power with one supply and open. Air handling unit fans and re: urn one exhaust fan inoperable. fans are operating.

Measure air flow to and from each diesel Recirculation duct isolanen d.urpers are generator bus switchgear room. closed.

'Ihe test shall be repeated with the other Supply and return branch duct isolation supply and exhaust fans inoperable. dampers are closed.

A minimum flow of [I2.(XK) sefm] is supplied and exhausted from each switchgear room.

WB5tillgh0llSe

Tier 1 Design Certification Material DIESEL GENERATOR BUILDING VENTILATION SYSTEM Revision: 1 Effective: 12/15/92 3,7.4 DIESEL GENERATOR BUILDING VENTILATION SYSTEM Design Description The diesel generator building ventilation system (VZS) serves no safety related function. This system provides ventilation, cooling, and heating of the diesel generator building for the onsite standby power system to perform its defense-in. depth functions.

The diesel generator building ventilation system has four subsystems. Two independent normal beating and ventilation systems, one per diesel generator, operate continuously. Two standby exhaust ventilation systems, one per diesel generator, operate in conjunction with diesel generator operation.

Each normal heating and ventilation subsystem consists of one primary air handling unit w hich performs the defense-in-depth function of ventilating the electrical equipment service module and me.intaining it below the design maximum temperature.

Each standby exhaust ventilation subsystem consists of exhaust fan (s) and associated outside air intake damper (s), which ventilate the diesel generator room and maintain it below the design maximum temperature when the diesel generator is operating.

3 4-'

w wesunsouse

Tier 1 Design Certification Material DIESEL GENERATOR BUILDING VENTILATION SYSTEM in r  !

Revision: 1 ,

Effective: 12/15/92 Table 3.7.4 Diesel Generator Staltling Ventilation System Inspections, Tests Analyses and Acceptance Criteria inspections, Tests, Analyses Acceptance Criteria Certified Design Commitment A test shall be performed by starting the a. He outside air intake damperts) epen, the

1. When the diesel generators are operating, a.

diesel generator. standby exhaust fan (s) start and the primary the diesel generator room and the electrical air handling unit operates.

equipmem sersice module are maintained at or below the design maximum temperature.

b. A test shall be conductM to operate the b. The diesel generator room temperature diesel generator at [2736] kw capacity for a does not exceed the outdoor temperature by time required to reach engine temperature more than [277 F.

equilibrium plus one hour.

He electrical equipment sersice compartment temperature does not exceed the outdoor temperature by more than

[4.5 fT.

3.7.4-2 W

Westinghouse 3

Ti:t 1 D:slon Certific: tion Mst: rial

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v NON SYSTEM TIER 1 MATERIAL Revision: 1 Ef f ective: 12/15/92

. i 4.0 NON SYSTEM TIER 1 MATERIAL Tier 1 entries are provided in this section for those subjects not conveniently addressed on a system basis, His includes:

• Human Itactors Enginecro.g

• Nuclear Island fluilding

• Safety-related Piping

• Interface.

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Ti:r 1 D: sign Certific: tion Mat:ri .I HUMAN FACTORS ENGINEERING V)

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Revision: 1 _!U Effective: 12/15/92 4.1 HUMAN FACTORS ENGINEERING Design Description The human factors engineering design includes the man-inachine design for the main control room and the remote shutdown workstation. The snain control room and the remote shutdown workstation conform to relevant human factors engineering guidelines. The num-machine interface in the main control room supports ef fective operation of the plant.

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G Tier 1 Design Certification Material HUMAN FACTORS ENGINEERING eij Revision: 1 Effective: 12/15/92 .

Table 4.1 Human Factors Engineering (Conformance to Human Factors Guidelines)

Inspections, Tests, Analyses and Acceptance Criteria' I

Certified Design Commitment inspections Tests. Analyses Acceptance Criteria j i

1. He main control room and the remote Inspections, tests or analyses shall be perform-d De evaluation of the human factors guidelines shutdown workstation conform to relevant to evaluate the design relevant to humanfactors exists and documents conformity to relevant  ;

human factors engineering guidelines. guidelines related to the following features: human factors guidelines.

• Environmental characteristics *

• Room arrangement

• Workstation (s) .

• Information display system (s)

• Alarm system (s)

• Controls

• Procedures

• Communication system (s) l i

f I

' 4.1-2 W westinghouse

x a Om Tier 1 Design Certification F.1aterial U

HUMAN FACTORS ENGINEERING lij Revision: 1 Effective: 12/15/92 Table 4.1 Human Factors Engneering (Conformance to Human Factors Guidelines) inspections, Tests, Analyses and Acceptance Criteria' Certified Design Commitment Inspections. Tests. Analyses Acceptance Criteria

1. (continued) Dese guidelin-s are derived frem the following references:

• NUREG41700, Guidelines for Control Room Design Reviews

• MIL-STD-1472

. Arnerican National Standard for Human Factors Engineering of Visual Display Terminals and Workstations

• ASHRAE Standard for nermal Comfort

. EPRI NP-3659. Human Factors Guide for Nuclear Power Plant Control Room Development.

' Design ITAAC 4.1-3 W Westinghouse

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kJ Tier 1 Design Certification Materia!

HUMAN FACTORS ENGINEERING -

Revision: 1 Effective: 12/15/92 _

Table 4.1 Human Factors Engineering (Validation of the Integrated M-MIS)

Inspections, Tests, Analyses and Acceptance Criteria

• Certified Design Commitment inspections, Tests, Analysis Acceptance Criteria

2. The man-machine interface in the main An analysis of the man-machine interface in the The results are documented and show that the control room supports effective operation of main control room shall be conducted using man-machine interface supports effective the plant. suitable simulation methoh and the results shall operation of the plant.

be documented in an evaluation report for a representative set of the foitowing plant operations:

• Startup a Shutdown

= Emergency operations.

2 Construction ITAAC W Westinghotise

I i

l Tier 1 Design Certific: tion M:teri:1 l l

NUCLEAR ISLAND BUILDINGS Revision: 1

)

Effective: 12/15/92 4.2 NUCLEAR ISLAND BUILDINGS Design Description Building Functions Containment Building The nuclear island buildings are the containment The containment building is the containment vessel building, shield building, and auxiliary building. The and the structures contained within the containment buildings ),rovide protection and support for safety- vessel. The containment buildingis a part of the overall related systems and components. The three buildings are containment system with the functions of containing the supported by a common basemat. The floor slabs and release of airborne radioactivity following postulated structural walls of the auxiliary building are structurally design basis accidents and providing shielding for the connected to the cylindricai section of the shield reactor core and the reactor coolant system during building. Le buildings are primarily reinforced normal operations. The containment vessel is a part of concrete and structural steel. Major structural elements the containment system and the passive containment of the nuclear island buildings meet the following cooling system. Refer to the containment system Tier requirements: 1 Design Description in Section 3.2.2 and to the passive containment cooling system Tier 1 Design Description

• Withstand the effects of postulated natural in Section 3.2.3. It is a freestanding cylindrical steel f

\

phenomena such as hurricanes, ikxxis. tornadns, vessel with elliptical upper and lower heads. Access to and earthquakes without loss of capability to the containment is provided through personnel airlocks perform safety-related functions. and equipment hatches.

• Withstand the effects of postulated internal events Shield Building such as fires, floodinF, and pipe rupture without loss of capability to perform safety-related The shield building is the shield building structure functions. and annulus area that surrounds the containment building. The shield building, in conjunction with the

• Provide radiation shielding to allow operator access internal structures of the containment building, provides to the main control room, and to limit radiation shielding for the reactor coolant system and the other exposure to safety-related systems and companents. radioactive systems and components housed in the containment. The shield building is a part of the passive The equipment in the nuclear island is arranged so containment cooling system. Refer to the passive that safe shutdown can be achieved for the postulated containment cooling system Tier 1 Design Description internal events. in Section 3.2.3.The shicM building raf supports the passive containment cooling system water storage tank and air diffuser. Air intakes are located along the cylindrical portion of the shield building.

4.2-1 b' W Westinghouse

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Tier 1 Design Certificction Materlil NUCLEAR ISLAND DUILDINGS jf lill Revision: 1 Effective: 12/15/92 Auxiliary Building The auxiliary building houses the safety related rnechanical and electrical equipment located outside the containment and shield buildings.The auxiliary building also provides 6hielding for the radioactive equipment and piping that is housed within the building. The auxiliary building is a C-shaped section of the nuclear island that wraps around a portion of the ihield building.

There are five floor levels in the auxiliary building (levels I through 5) with the lowest level designated as level 1.

The auxiliary building is divided into radiologically controlled areas and nonradiologically controlled areas which are physically separated by structural walls and floor slabs. These structural barriers are designed to prevent flood and/or fire propagation across the boundary between these areas. The building is further subdivided into fire areas separateil by fire rated structural barriers, W WB511nghouse h

w Tier 1 Design Certification Material NUCLEAR ISLAND BUILDINGS iEE: EE m =

Revision: 1 --

Effective: 12/15/92 ...

Table 4.2 Nuclear Island Buildings I inspections. Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. The nuclear island buildings are designed for Inspections and/or analyses shall be A report is available to confirm that the btuldings normal loads, extulated natural phenomena performed of the as-built maior structural are designed for the rmrmal hwis, pcwtulated and postulated intemal events. elements of the nuclear island buildings. natural phenomena and pwtulated internal Results shall be documented in a rep >rt. events.

2. The exterior walls of the nuclear island, a) Inspections shall be performed to serify a) Seals are installed at vmgs through the including any penetratmns, are designed to that openings through the exterior walls exterior wa'Is below the extemal thxw! level.

protect against extemal flmsl. beknv the external ihmd level incorporate seals to prevent flomling.

b) Omstruction dxuments shall be b) Waterproofing is installed en the outside of reviewed to venfy that waterproofing is the exterior walls below the extsmal finxi installed on the outside of the extemal level.

walls below the extemal flom! level.

3. Design features are provided to protect the Analyses shall be ctmducted for potential A report is available to confirm tnat design ajuipment required for safe shut kum against sources of intemal flomling to demonstrate features are pmvidal to protect the equipment intemal flaxfing. that safe shutdown can be achieved; and a required for safe shut &mn against intemal vicual inspection shall be conducted of as- flawling.

built design features (drains, curbs, etc.)

provided to prevent flooding of equipment required for safe shutdown. Results shall be docummted in a rep >rt.

W-Westingflouse

Q ~.J V)

Tier 1 Design Certification Material 3E =;;-

NUCLEAR ISLAND BUILDINGS -

'E Revision: 1 .

Effective: 12/15/92 _

Table 4.2 Nuclear Island Buildings inspections, Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria

4. Design features are provided to protect Analyses shall be conducted for potential A report is available to confirm that design i fires in each area of the nuclers island to features are provided to protect the equipment equipment required for safe shutdown against fire. demonstrate that safe shutdown can be required for safe shutdown against fire.

achieved; and a visual inspection shall be conducted of design features (fire barriers.

fire doors, penetration seals, etc.) provided to protect equiptrent required for safe shutdown. Results shall be documented in a report Design features are providcd to protect Analyses of the effects of postulated pipe A reptnt is available to confirm that design 5.

ruptures shall be performed to demonstrate features are provided to protect the equipment equipment required for safe shutdown against that safe shutdown can be achieved. Design required for safe shutdown against postulated the postulated rupture of high and rnoderate energy fluid systems. against dynamic effects may be excluded for pipe ruptures.

those piping systems that are qualified for the optional leak-before-break design t

approach. A visual inspection shall be I

conducted of the as-built piping systems and the design features provided to protect equipment required for safe shutdown against postulated pipe ruptures.

W Westinghouse

Tier 1 Design Certification Materlil n

SAFETY-RELATED PIPING (v) Revision: 1 i!!!

Effective: 12/15/92 4.3 SAFETY RELATED PIPING Desi0n Description Piping is categorized based on the functions of the system. Design requirements are established far ech safety related piping system and interfaces with Ltructures, systems and Components are identified.

Stress analyses demonstrate that the as-built piping satisfies the specified requirements and interfaces.

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O O Tier 1 Design Certificat on Material i

O SAFETY-RELATED PIPING ls= at E

Revision: 1 Effective: 12115/92 Table 4.3 Safety-related Piping Inspections Tests, Analyses and Acceptance Criteria Certified Design Commitment inspections, Tests, Analyses Acceptance Criteria I. Design specifications provide a basis for the Inspection (review) of the design specification ne existence of a design vecification is construction of the safety-related piping and shall be conducted for the safety-relatal pntions cutirmed for the safety-related portions of the shall include the folkming: of the systems listed in Table 4.3-2. systems listed in Table 4.3-2.

• He functions and boundaries of the safety-related piping

• The design and service conditions including cyclic loads for 60 years for the reactor coolant pressure boundary

• The environmental conditions '

= ne material requirements

= The metixxis for the dynamic and static analysis of piping systems

• The functional capability requirements

• The pressure boundary integrity requirements.

West l11gflouse

O O O Tier 1 Design Certification Materid pip SAFETY-REl.ATED PIPING h Revision: 1 .

Effective: 12/15/92 I

Table 4.2 Safety-related Piping Inspections, Tests. Analyses and Acceptance Criteria inspections. Tests, Analyses Acceptance Criteria Certified Design Commitment i

The N Symbol Stamp is confirmed for the

2. Safety-related piping meets the design The pipe routing; the location, orientation, and size of snubbers and struts; the kation and size safety- related portions of the systems listed in specification. The loads, accelerations, and Table 4.3-2.

stresses that the piping system imposes on its of hangers; the location and weight of valves, pumps, and hest exchangers; the location and pipe mounted equipment and on its interfaces with structures and other components and configuration of anchors; the location of guides and pipe whip restraints; and the specified piping is compared to the allowable values.

clearances, shall be confirmed by resiewing Functionalinterference is avoided with other drawings, and by performing a visualinspection piping, structures, and components as the of the installed piping. The as-built information piping moves or deflects due to the thermal, shall be reviewed in conjunction with the as-dynamic, and/or static loads which it analyzed piping system. The design report shall experiences in service.

be inwted.

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W westinghouse

Tier 1 Design Certification Material SAFETY RELATED PIPING g ,

Revision: 1 T ,

Effective: 12/15/92 .

e Table 4.3 2 Systems including Safety Related Piping Tier 1 Design System Description Section 3.1.2 Reactor coolant system 3.2.1 Automatic depressuriution system 3.2.2 Containment system (applicable to piping connected to containment piping penetrations, refer to Table 3.2.2-2) 3.2.3 Passive containment cooling system 3.2.4 Passive core cooling system 3.2.5 Steam generator system 3.2.6 Mam control room emergency habitability system 3.3.2 Chemical and volume control system 3.3.6 Primary sampling system 3.3.7 Normal residu4l heat removal system 3.3.8 Spent fuel pit cooling system W Westinghouse

Tier 1 D: sign Certific: tion M t: rial INTERFACE  ;

'v Revision: 1 i l

Effective: 12/15/92 e

4.4 INTERFACE 10 ClR 52.47 (a)(1)(viii) requires that compliance with the interf ace requirements to be met by those portions of the plant for which the application does not seek certification be verifiable through inspection, testing (either in the plant or elsewhere), or analysis. The Al%00 is a plant design incorporating the entire nuclear island, the annen buildings and associated equipment, the diesel / generator building and associated equipment, the turbine generator building, the turbine / generator equipment and the radwaste facilities. As a result, no interfaces need to be identified between or among these portions of the plant.

There are safety related or defense-in-depth interfaces between the Al%00 design and other portions of a facility having a combined license under 10 CFR Part 52 which must be addressed by parties that reference the Al%00 design. The interfaces between the AP600 p safety-related and defense-in-depth systems and the other

( portions of such a facility are addressed in the specific systents ITAAC.

The AP600 site parameters are described in Section 5.0.

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SITE PAFIAMETERS Revision: 1 l$!

Effective: 12/15/92 H 5.0 SITE PARAMETERS Design Description ne AP600 is designed for site parameters that envelope the conditions that will occur at rnost potential power plant sites in the United States. Table 5.0-1 identifies the key site parameters that are specified for the design of safety-related aspects of structures, systems and components.

Table 5.0-1 Sito Parameters Air Temperature Limits based on historical data excluding peaks of less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> duration.

Maximum dry bulb temperature of Il5'P Minimum dry bulb te:nperature of -40'F Tornado Wind Speed Maximum wind speed of 300 rnph Safe Shutdown Earthquake (SSE) SSE free field peak ground acceleration of 0.30g with llorimntal and Vertical Ground Response Spectra as given in Figures 5.0-1 and 5.0-2 Precipitation Rain 19.4 in./hr (6.2 in./5 min)

Snow /lce Ground Snow Load of 50 pounds per square foot O

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[\ 0 V 'V Tier 1 Design Certification Material SITE PARAMETERS t; =is

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Revision: 1 E Effective: 12/15/92 _

R.G. 1.60_Horarontal Desagn Response Spectra 2.0 , ,

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! Figure 5.0-1 Safe Shut &mn Earthquake Horizontal Design Response Spectra 5.0-2 l W Westinghouse

O O Tier 1 Design Certification Material O

- SITE PARAMETERS =g Revision: 1 Effective: 12115/92 RG 1.60 Vertical Design Response _ Spectre 2.0 , ,

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Figure 5.0-2 Safe Shutdown Eanhquake Vertical Design Response Spectra

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